ML15222A846

ENT000635 - Westinghouse, WCAP-17096-NP, Rev. 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements (Dec. 2009)

ML15222A846
Person / Time
Site:	Indian Point
Issue date:	12/31/2009
From:	Entergy Nuclear Operations
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 28134, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15222A846 (212)

Text

ENT000635 Submitted: August 10, 2015

- .... - - -.., . .1

. "I'll 2ý . . I - ,_ _ --.- ý - -_

Westinghouse Non-Proprietary Class 3 WCAP-1 7096-NP December 20109 Revision 2 Reactor Internals Acceptance Criteria Methodology and Data Requirements Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17096-NP Revision 2 Reactor Internals Acceptance Criteria Methodology and Data Requirements Randy Lott*

Reactor Internals Design and Analysis Steve Fyfitch Materials and Structural Analysis (AREVA NP Inc.)

December 2009 Reviewer: Joshua McKinley*

Reactor Internals Design and Analysis Approved: Brian Gaia*, Manager Reactor Internals Design and Analysis Work Performed under PA-MSC-0473

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

© 2009 Westinghouse Electric Company LLC All Rights Reserved WCAP-17096-NP.doc-120209

111 LEGAL NOTICE This report was prepared as an account of work performed by Westinghouse Electric Company LLC. Neither Westinghouse Electric Company LLC, nor any person acting on its behalf:

A. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report.

COPYRIGHT NOTICE This report has been prepared by Westinghouse Electric Company LLC and bears a Westinghouse Electric Company copyright notice. As a member of the PWR Owners Group, you are permitted to copy and redistribute all or portions of the report within your organization; however all copies made by you must include the copyright notice in all instances.

DISTRIBUTION NOTICE This report was prepared for the PWR Owners Group. This Distribution Notice is intended to establish guidance for access to this information. This report (including proprietary and nonproprietary versions) is not to be provided to any individual or organization outside of the PWR Owners Group program participants without prior written approval of the PWR Owners Group Program Management Office. However, prior written approval is not required for program participants to provide copies of Class 3 Non Proprietary reports to third parties that are supporting implementation at their plant, and for submittals to the NRC.

WCAP-17096-NP December 2009 Revision 2

iv PWR Owners Group Member Participation* for PA-MSC-0473 Utility Member Plant Site(s) Participant Yes No AmerenUE Callaway (W) X American Electric Power D.C. Cook l&2 (W) X Arizona Public Service Palo Verde Unit 1, 2, & 3 (CE) X Constellation Energy Group Calvert Cliffs 1 & 2 (CE) X Constellation Energy Group Ginna (W) X Dominion Connecticut Millstone 2 (CE) X Dominion Connecticut Millstone 3 (W) X Dominion Kewaunee Kewaunee (W) X Dominion VA North Anna I & 2, Surry I & 2 (W) X Duke Energy Catawba 1 & 2, McGuire 1 & 2 (W) X Oconee 1, 2, 3 (B&W) X Entergy Palisades (CE) X Entergy Nuclear Northeast Indian Point 2 & 3 (W) X Entergy Operations South Arkansas 2, Waterford 3 (CE) X Arkansas 1 (B&W) X Exelon Generation Co. LLC Braidwood 1 & 2, Byron 1 & 2 (W) X TMI 1 (B&W) X FirstEnergy Nuclear Operating Co Beaver Valley 1 & 2 (W) X Davis-Besse (B&W) X Florida Power & Light Group St. Lucie 1 & 2 (CE) X Turkey Point 3 & 4, Seabrook (W) X Pt. Beach l&2 (W) X Luminant Power Comanche Peak 1 & 2 (W) X Xcel Energy Prairie Island 1&2 X Omaha Public Power District Fort Calhoun (CE) X Pacific Gas & Electric Diablo Canyon 1 & 2 (W) X Progress Energy Robinson 2, Shearon Harris (W) X Crystal River 3 (B&W) X PSEG - Nuclear Salem 1 & 2 (W) X Southern California Edison SONGS 2 & 3 (CE) X South Carolina Electric & Gas V.C. Summer (W) X So. Texas Project Nuclear Operating Co. South Texas Project 1 & 2 (W) X Southern Nuclear Operating Co. Farley 1 & 2, Vogtle 1 & 2 (W) X Tennessee Valley Authority Sequoyah 1 & 2, Watts Bar (W) X Wolf Creek Nuclear Operating Co. Wolf Creek (W) X Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above.

WCAP- 17096-NP December 2009 Revision 2

V PWR Owners Group International Member Participation* for PA-MSC-0473 Utility Member Plant Site(s) Participant Yes No British Energy Sizewell B X Electrabel (Belgian Utilities) Doel 1, 2 & 4, Tihange 1 & 3 X Hokkaido Tomari 1 & 2 (MHI) X Japan Atomic Power Company Tsuruga 2 (MHI) X Mihama 1, 2 & 3, Ohi 1, 2, 3 & 4, X Takahama 1, 2, 3 &4 (W & MHI)

Kori 1, 2,3 & 4 X Korea Hydro & Nuclear Power Corp. Yonggwang 1 & 2 (W)

Korea Hydro & Nuclear Power Corp. Yonggwang 3, 4, 5 & 6 X Ulchin 3, 4,5 & 6(CE)

Kyushu Genkai 1, 2, 3 & 4, Sendai 1 & 2 (MHI) X Nukleama Electrarna KRSKO Krsko (W) X Nordostschweizerische Kraftwerke AG Beznau 1 & 2 (W) X (NOK)

Ringhals AB Ringhals 2, 3 & 4 (W) X Shikoku Ikata 1, 2 &.3 (MHI) X Spanish Utilities Asco 1 & 2, Vandellos 2, X Almaraz 1 & 2 (W)

• Taiwan Power Co. Maanshan 1 & 2 (W) X Electricite de France 54 Units X

• This is a list of participants in this project as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending documents to participants not listed above.

WCAP- 17096-NP December 2009 Revision 2

vi TABLE OF CONTENTS L IST O F TAB L E S ... .................................................................................................................................... ix L IST O F FIG UR E S ..................................................................................................................................... xi LIST OF ACRONYMS AND ABBREVIATIONS .................................................................................... xiii A CKN O WLED G EMEN T S ....................................................................................................................... 1-1 S OB JEC T IV E ................................................................................................................................. 1-2 2 B A C K GR OU N D .......................................................................................................................... 2-1 3 ANA LY SIS PRO C E SS ................................................................................................................ 3-1 3.1 PHYSICAL MEASUREMENTS .................................................................................... 3-1 3.2 G ENER AL CO ND ITION ............................................................................................... 3-1 3.3 ULTRASONIC TESTING (BOLTS) ............................................................................... 3-2 3.4 V ISUAL C RACK IN G ..................................................................................................... 3-2 3.5 V ISUAL OT H E R ............................................................................................................. 3-3 3.6 FATIGUE (QUALIFY BY TIME-LIMITED AGING ANALYSIS [TLAA]) ................. 3-3 3.7 RESOLUTION BY AN ALY SIS ...................................................................................... 3-3 4 RECOMMENDATION FORMAT ............................................................................................... 4-1 4.1 IDENTIFICATION AND EXAMINATION RECOMMENDATIONS ........................... 4-1 4.2 COM PONEN T FUN CTION ........................................................................................... 4-1 4.3 IN SPECTION OU TCO M ES ........................................................................................... 4-1 4.4 METHODOLOGY AND DATA REQUIREMENTS ...................................................... 4-1 4.5 RECOMMENDED APPROACH .................................................................................... 4-2 5 B& W PLANT DESIGN RESULTS .............................................................................................. 5-1 5.1 M ETH O D O L O G IE S ....................................................................................................... 5-1 5.2 GENERIC ACCEPTANCE CRITERIA ................................. 5-1 6 COMBUSTION ENGINEERING AND WESTINGHOUSE PLANT DESIGN RESULTS ....... 6-1 6.1 M E THO D O LO G IE S ....................................................................................................... 6-1 6.2 GENERIC ACCEPTANCE CRITERIA .......................................................................... 6-1 7 REFEREN C E S ............................................................................................................................. 7-1 APPENDIX A B&W DESIGN PRIMARY AND EXPANSION COMPONENT ITEM ACCEPTANCE CRITERIA METHODOLOGY AND DATA REQUIREMENTS ............... A-1 APPENDIX B ACCEPTANCE CRITERIA FLOWCHARTS FOR B&W-DESIGNED COMPONENTS INCLUDED IN MRP-227 ......................................................................... B-1 WCAP- 17096-NP December 2009 Revision 2

vii APPENDIX C ACCEPTANCE CRITERIA METHODOLOGY AND DATA REQUIREMENTS FOR COMBUSTION ENGINEERING COMPONENTS INCLUDED IN MRP-227 ........... C-1 APPENDIX D FLOW CHARTS OF ILLUSTRATING EVALUATION METHODOLOGIES FOR COMBUSTION ENGINEERING-DESIGNED PLANTS ............................................ D-l APPENDIX E ACCEPTANCE CRITERIA METHODOLOGY AND DATA REQUIREMENTS FOR WESTINGHOUSE COMPONENTS INCLUDED IN MRP-227 ................................. E-1 APPENDIX F FLOW CHARTS OF ILLUSTRATING EVALUATION METHODOLOGIES FOR WESTINGHOUSE-DESIGNED PLANTS ................................................................... F-1 WCAP-17096-NP December 2009 Revision 2

ix LIST OF TABLES Table 5-1 Applicability of Potential Generic Acceptance Criteria for B&W-Design Primary and Expansion Component Item s .................................................................................... 5-2 Table 6-1 Applicability of Potential Generic Acceptance Criteria for CE-Design Primary and Expansion Com ponent Item s .................................................................................... 6-2 Table 6-2 Applicability of Potential Generic Acceptance Criteria for Westinghouse-Design Primary and Expansion Component Items ...................................................................... 6-4 WCAP- 17096-NP December 2009 Revision 2

xi LIST OF FIGURES None.

December 2009 WCAY- 17096-NP WCAP-17096-NP December 2009 Revision 2

xiii LIST OF ACRONYMS AND ABBREVIATIONS ANO Arkansas Nuclear One ASME American Society of Mechanical Engineers B&W Babcock & Wilcox B&WOG Babcock & Wilcox Owners Group BB baffle-to-baffle BF baffle-to-former BMI bottom-mounted instrumentation BWR boiling water reactor BWRVIP boiling water reactor vessel internals program CASS cast austenitic stainless steel CE Combustion Engineering CEA control element assembly CF core barrel-to-former CGR crack growth rate CLB current licensing basis COD crack opening displacement CR Crystal River CRGT control rod guide tube CSS core support shield CW cold worked DB Davis-Besse DNB departure from nucleate boiling EFPY effective full-power year EPRI Electric Power Research Institute ET eddy-current test FD flow distributor FEA finite element analysis FEM finite element model FMEA failure modes and effects analysis HAZ heat-affected zone HWC hydrogen water chemistry I&E inspection and evaluation IASCC irradiation-assisted stress corrosion cracking IC irradiation creep ID inside diameter IE irradiation embrittlement IGSCC intergranular stress corrosion cracking IMI incore monitoring instrumentation ISI in-service inspection ISR irradiation-induced stress relaxation JCO justification for continued operation LCB lower core barrel LEFM linear-elastic fracture mechanics LOCA loss-of-coolant accident WCAP- 17096-NP December 2009 Revision 2

xiv LIST OF ACRONYMS AND ABBREVIATIONS (cont.)

LTS lower thermal shield MRP Materials Reliability Program NDE nondestructive examination NRC U.S. Nuclear Regulatory Commission NSSS nuclear steam supply system OD outside diameter OE operating experience ONS Oconee Nuclear Station PWR pressurized water reactor PWROG Pressurized Water Reactor Owners Group PWSCC primary water stress corrosion cracking RICT Reactor Internals Core Team RIFG Reactor Internals Focus Group SCC stress corrosion cracking SCF stress concentration factor SSE safe-shutdown earthquake SSHT surveillance specimen holder tube TAC Technical Assignment Control TBD to be determined TE thermal embrittlement TLAA time-limited aging analysis TMI Three Mile Island UCB upper core barrel UT ultrasonic test UTS upper thermal shield WOG Westinghouse Owners Group Trademark Statement:

WINCISE is a trademark of Westinghouse Electric Company LLC.

WCAP- 17096-NP December 2009 Revision 2

1-1 ACKNOWLEDGEMENTS The authors would like to thank Jim Cirilli, Mike McDevitt, and the members of the PWROG Materials Subcommittee for their review and comments on Revision 0 of this WCAP. The authors would also like to acknowledge the contributions of their colleagues at AREVA-NP and Westinghouse in developing this document.

WCAP- 17096-NP December 2009 Revision 2

1-2 1 OBJECTIVE The objective of this project is to identify consistent, industry-wide analytical methodologies and data requirements for developing:

1. Acceptance Criteria for the Primary and Expansion Components identified in the Materials Reliability Program (MRP) Reactor Internals Inspection and Evaluation (I&E) Guidelines (MRP-227 Rev. 0)

2. Evaluation Procedures for utilities to assess potential safety and functional impacts of degradation in components with observed relevant conditions These criteria and procedures must be established and generally accepted across the industry prior to the implementation of the I&E Guidelines. This effort supports the I&E recommendations of MRP-227. It is anticipated that the methodologies and data requirements defined in this effort will be reviewed by the U.S. Nuclear Regulatory Commission (NRC) in the course of the MRP-227 Safety Evaluation.

The current project is only Phase I of the efforts required to develop acceptance criteria. The potential for generic efforts for each component is also evaluated in this program, and where practicable, work scope for a follow-on program to address the generic analyses of these components is identified. Because the component lists are specific to the original nuclear steam supply system (NSSS) suppliers, it is anticipated that any effort to develop acceptance criteria for specific components will be submitted to the Pressurized Water Reactor Owners Group (PWROG) for cafeteria funding.

For each of the Primary and Expansion Components listed in MRP-227, this report outlines:

• Type of analyses required 0 Required evaluation procedures

• Data required to support analysis 0 Logic chart illustrating evaluation path and potential disposition options

• Component items (Primary and Expansion) that can be addressed on a generic basis Note that letter OG-09-290 1 issued Revision 0 of this document for review and comment by the PWROG Materials Subcommittee on July 27, 2009. Comments from this review have been incorporated in the current version.

1 Letter OG-09-290, "Transmittal of Draft Report WCAP- 17096, Rev 0 "Reactor Intemals Acceptance Criteria Methodology and Data Requirements", PAMSC-0473," July 27, 2009.

WCAP- 17096-NP December 2009 Revision 2

2-1 2 BACKGROUND The MRP Reactor Internals Focus Group (MRP RIFG) issued a formal request for the PWROG to sponsor a project to develop "generic" acceptance criteria for the proposed MRP-227 I&E Guidelines [1]

on March 19, 2008 (Reference MRP Letter 2008-026). The MRP RIFG submitted the I&E Guidelines to the NRC for a formal Safety Evaluation in December 2008. At that time, the Guidelines would be submitted under NEI 03-08 Procedures. The schedule of this project was adopted to match the requirements for NRC review.

The I&E Guidelines in [1] for pressurized water reactor (PWR) internals are based on a thorough screening of both potential degradation and operating experience [2, 3, 4]. They are designed to target inspections of locations where aging degradation can potentially impair component function.

Functionality analyses associated with the original screening evaluations have identified the potential operational concerns, and inspection methodologies have been identified for each component [5, 6].

However, the current I&E Guidelines do not provide detailed acceptance and evaluation criteria for each component.

The I&E Guidelines in MRP-227 build on the existing ASME Code,Section XI inspections to create comprehensive inspection recommendations for aging degradation in reactor internals [7,8]. The fundamental goal of any inspection under the ASME Code,Section XI program [9] is to identify relevant conditions that require further action. Article IWA-9000 of the ASME Code defines relevant condition as follows:

Relevant Condition - A condition observed during a visual examination that requires supplemental examination, corrective measure, correction by repair/replacement activities, or analytical evaluation.2 The ASME Code [9] defines acceptance standards in IWB-3400, which are used to determine whether an observed condition is acceptable for service or is a relevant condition. The inspection standards used to define relevant condition are based on generic analysis that provides a high level of assurance of satisfactory function. Acceptance standards for ASME Examination Categories B-N-2 (Welded Core Support Structures and Interior Attachments to Reactor Vessels) and B-N-3 (Removable Core Support Structures) are provided in IWB-3520. While these acceptance standards are appropriate for the ASME Code Section XI inspections, including those that are specifically highlighted in MRP-227 as existing reactor internals inspection requirements, they are not fully applicable to the MRP-227 new inspection recommendations. In particular, the referenced linear flaw standards of IWB-3510 are intended to guard against propagation of cracks through the reactor pressure vessel are not meaningful when applied to the removable internals components.

2. ASME (Article IWA-9000) uses this definition for visual examinations. The definition has been expanded here to include all inspections conducted under the I&E Guidelines.

WCAP- 17096-NP December 2009 Revision 2

2-2 MRP-227 provides lists of components with specific inspection recommendations. The components are divided into three basic categories based on the recommendation:

Primary Components - Inspection of the Primary Components is required in the I&E Guidelines.

In general, these inspections must be conducted early in the license extension period. Every plant needs to have acceptance criteria for Primary Components included in their aging management program.

Expansion Components - The MRP -227 defines criteria for the results of Primary Component inspections that will trigger inspection of the Expansion Components. Although the MRP-227 requirements allow a time delay for mobilization of equipment and resources, acceptance criteria for the expansion components will also be required.

Existing Components - There are established acceptance criteria for the ASME Code,Section XI and other exams in this category. The PWROG may want to review the current field practice on these inspections as part of the implementation evaluations to ensure a consistent approach.

Each of these components also requires some acceptance standard for evaluating the relevant condition.

For many general condition monitoring visual examinations (VT-3), MRP-227 identifies specific conditions that supplement the list of relevant conditions provided in IWB-3520.2. However, there is no comprehensive listing of acceptance standards in MRP-227. The general standards for determining relevant conditions in reactor internals inspections are provided in MRP-228 [10]:

Welds - Cracks, or indications that exhibit characteristics of cracking, are considered relevant.

Components - Cracking or other significant degradation that could impair the ability of the component to perform its design function is considered relevant.

Once a relevant condition has been identified, evaluation is required to assess the ability of the degraded component to continue to perform the design function without interfering with the function of the system.

The evaluation of suitability for continued service will be, in general, an extension of the analysis used to define relevant condition. Although specific acceptance criteria and evaluation procedures need not be fully developed at the time that the I&E Guidelines are submitted to the NRC, it is critical that there be a clear path to successful implementation. The scope of this PWROG Reactor Internals Core Team (RICT)

Project Authorization is to provide this path by defining the process for developing these acceptance criteria and evaluation procedures. The process will be defined in the following terms:

Evaluation Methodology - The procedures and criteria to be used by the engineering staff to evaluate relevant conditions. This includes:

- Demonstration of functionality of the current configuration

- Establishment of a re-inspection frequency of one or more refueling cycles S- An engineering basis for repair/replacement/mitigation options WCAP- 17096-NP December 2009 Revision 2

2-3 Acceptance Criteria - The criteria against which the need for corrective action will be evaluated.

The acceptance criteria should ensure that the intended functions of the particular structure and component are maintained under all current licensing basis design conditions during the period of 3

extended operation.

In some cases, it will be feasible to avoid plant-specific evaluations by adopting generic standards for acceptance that will ensure compliance with the accepted industry requirements.

Generic Acceptance - Disposition of a relevant condition based on a generic implementation of the evaluation methodology for an NSSS design or other plant grouping.

The original MRP request forwarded to the PWROG RI-CT was for generic acceptance criteria.

However, as a prerequisite to developing the generic criteria, the evaluation methodology must be defined. The development of evaluation methodologies also poses significant implementation issues that need to be considered by the PWROG. Therefore, the Phase I program summarized in this document outlines the evaluation methodology for components to be inspected under the proposed I&E Guidelines and identifies analyses that may be conducted on a generic basis. The Phase I program defines potential scope for additional PWROG projects to support the development of evaluation methodologies and generic acceptance criteria.

3. Definition based on [12] Aging Management Program (AMP), Element 6.

WCAP- 17096-NP December 2009 Revision 2

3-1 3 ANALYSIS PROCESS The goal of this effort is to define the process to be used in the degradation evaluation procedures and the necessary data requirements to perform the evaluation of each Primary and Expansion component such that engineering organizations follow a consistent approach that is documented and approved. The proposed guidelines and methodologies will be listed in outline format and illustrated in a logic chart showing potential evaluation and disposition options for the relevant condition.

Both AREVA and Westinghouse assembled teams of experts to develop recommendations. Although the procedural details and reporting format of these efforts are unique to the vendors, the basic process followed the same general steps:

1. Review MRP-227 degradation modes and inspection recommendations.

2. Determine component function.

3. List potential inspection outcomes and observable effects.

4. Identify potential failure mechanisms and effects.

5, Outline methodology to evaluate potential inspection observations.

6. Determine data requirements for inspection.

7T Consider potential for vendor-specific generic analysis.

The form and structure of the evaluation methodology is determined by the type of inspection recommended.

3.1 PHYSICAL MEASUREMENTS Physical measurements are employed to characterize changes in component dimension. For example, measurements of internals hold-down spring height are used to evaluate loss of core hold-down forces due to wear or stress relaxation. In this case, the required hold-down forces are a design requirement and generic or plant-specific acceptance criteria should be established prior to the inspection.

No specific action to establish general acceptance criteria was required under this task. The required methodologies will simply note the existence of relevant design requirements for the affected components.

3.2 GENERAL CONDITION There is a heavy reliance on general condition monitoring within the I&E recommendations. Although these recommendations refer to the VT-3 level visual inspections, the guidelines generally provide a description of the expected degradation. Under the VT-3 procedure, any observation of degradation is considered to be a relevant condition and is reported to engineering for evaluation. Generally, engineering will review the observation to either confirm the relevant condition or disposition it as a visual anomaly.

Due to the qualitative nature of the VT-3 examination, it is difficult to define quantitative evaluation and acceptance criteria. In some cases, VT-3 examinations have been specified for redundant components, where multiple failures are required to impair the functionality of the system. In other cases, VT-3 has WCAP- 17096-NP December 2009 Revision 2

3-2 been specified where the first signs of degradation are expected to be visual (e.g., wear). In any case, it is prudent engineering practice to anticipate the range of possible visual observations and define resolution strategies.

A process identifying potential relevant conditions arising from each recommended VT-3 examination and defining resolution strategies is required for each VT-3 examination. This process may take the form of a failure modes and effects analysis (FMEA).

3.3 ULTRASONIC TESTING (BOLTS)

Within the I&E Guidelines, ultrasonic testing (UT) examinations are conducted to identify failed bolts.

All bolts with positive indications of cracking are assumed to be failed. The bolting systems in the internals are generally highly redundant. Acceptance criteria are based on minimum bolting patterns that guarantee structural stability through both normal operation and design basis transients. To establish an appropriate inspection interval, the current distribution of unfailed bolts must contain sufficient margin to demonstrate that the number of anticipated failures will not cause the distribution to fall below the minimum pattern. This will require either a historical analysis of bolt failure rates or a detailed model of bolt failure mechanisms.

The use of minimum bolting patterns as acceptance criteria that allow individual bolt failures is established in the industry. The PWROG (from prior Westinghouse Owners Group [WOG] efforts) has developed minimum bolting patterns for baffle-to-former bolting for Westinghouse reactor internals designs. The PWROG has supported development of similar strategies for core barrel bolt inspections in Babcock & Wilcox (B&W) plants. The methodology for performing a minimum bolting pattern or similar strategy is beyond the scope of the current task.

3.4 VISUAL CRACKING Visual examinations to identify cracking are generally recommended where intergranular stress corrosion cracking (IASCC), stress corrosion cracking (SCC), or fatigue is identified as the cracking mechanism.

The initial visual examinations in the B&W plants are all based on VT-3 requirements. The VT-3 exams were deemed to be adequate because the structures are relatively flaw tolerant. Appropriate follow-on actions might include EVT-1, UT, or eddy-current testing (ET) exams to determine flaw size. These options would be identified as part of the recommended procedure for resolution of the original VT-3 observation. Consistent with the current state-of-the-art within the Boiling Water Reactor Vessel and Internals Project (BWRVIP), EVT- 1 has been identified as the appropriate visual examination procedure for several of the Combustion Engineering (CE) and Westinghouse components. The EVT-1 examination can produce flaw size information that would lead directly to a fracture mechanics evaluation. UT or ET may also be used as enhanced or supplemental examinations, where appropriate.

General fracture mechanics procedures for calculating critical flaw sizes and growth rates are described within the I&E Guidelines. In order to apply these procedures, the appropriate irradiation history, loading conditions and stress intensity solutions must be identified. These factors are all dependent on both the flaw location and the plant design.

WCAP- 17096-NP December 2009 Revision 2

3-3 This task outlines specific fracture mechanics analysis requirements for each of the visual EVT-1 examinations included in the Primary and Expansion tables of the I&E Guidelines.

3.5 VISUAL OTHER A VT- 1 level examination was identified to examine potential swelling-related distortion in some welded core shroud structures originally designed by CE. The intention of the examination is to provide semi-quantitative data that can be used to evaluate the overall level of swelling in the structure. The original functionality analysis, which is known to be conservative, predicted large gap openings at specific locations in the shroud structure.

These examinations are meant to provide an early warning of swelling in the structure. Evaluation of this swelling-related data will require additional sensitivity studies to relate the swelling level to the predicted distortion and gap opening in the structure. Sensitivity studies are currently being considered by the MRP and are beyond the scope of this project.

3.6 FATIGUE (QUALIFY BY TIME-LIMITED AGING ANALYSIS [TLAAI)

Four component items in the CE design are included in the list of Primary components solely due to concerns about fatigue. Due to the plant-specific nature of the TLAA required for license extension programs, fatigue analysis was not included in the MRP functionality analysis. It is considered a high probability that TLAA will demonstrate a negligible probability of fatigue crack initiation in these components. However, pending resolution by TLAA, these component items are included in the Primary Component list. Acceptance criteria for these four fatigue-related items are not included in this task.

3.7 RESOLUTION BY ANALYSIS Three Expansion component items in the B&W designs have been designated for resolution by analysis.

Inspection of these three components is considered to be impractical due to issues of accessibility.

Therefore, should concerns about the integrity of these components be triggered by observations in the associated Primary components, no inspection is required. Resolution would require either detailed analysis or replacement.

Acceptance criteria for these three Expansion components are not included in this task. A separate project authorization to support additional analysis of these three B&W components may be proposed to the PWROG.

WCAP- 17096-NP December 2009 Revision 2

4-1 4 RECOMMENDATION FORMAT Recommendations for each Primary and Expansion item were prepared by Westinghouse and AREVA after consulting with appropriate experts and expert panels. The basic requirements for this study were mutually agreed upon between the vendors and defined in the original project authorization. The internal processes employed to complete these studies were specific to the vendor. The results of the Westinghouse and AREVA analyses are provided in Appendix A (B&W plants), Appendix C (CE plants),

and E (Westinghouse plants). Although there are superficial differences in the format in which the results are presented by each vendor, the underlying structure of the information is the same.

4.1 IDENTIFICATION AND EXAMINATION RECOMMENDATIONS Component identification and inspection recommendations for both the Primary and Expansion component items were originally provided in MRP-227. Relevant rows from MRP-227 Tables 4.1 and 4.4 were reproduced as part of the AREVA analysis of component items from the B&W plants as shown in Appendix A. Corresponding information on CE plants was extracted from MRP-227 Tables 4.2 and 4.5 and integrated into the Westinghouse data forms shown in Appendix C. A similar process was used to extract information on Westinghouse plants from MRP-227 Tables 4.3 and 4.6 for the data for the Westinghouse data forms in Appendix E.

4.2 COMPONENT FUNCTION To facilitate consideration of consequence of failure, both vendor processes required a brief summary of the component function. More extensive function descriptions for the reactor internals components were originally compiled in support of the Consequence Analysis performed for the Issue Management Tables.

These summaries are included in MRP- 156 [11]. The component function summaries provided in Appendices A, C, and E are meant only to provide perspective for the analysis recommendations.

4.3 INSPECTION OUTCOMES The inspection recommendations provided in MRP-227 are based on the identification of specific aging degradation mechanisms. The expert teams were asked to consider the information potentially generated by the inspection. The AREVA evaluations are summarized in Appendix A under the headings "Observable Effects" and "Possible Outcomes." The Westinghouse evaluations are summarized in Appendices C and E under the headings "Observable Effect," "Failure Mechanism," "Failure Effect," and "Failure Criteria."

4.4 METHODOLOGY AND DATA REQUIREMENTS The objective of this study was to define methodologies and data requirements for analysis and acceptance of degraded components identified in the MRP-227 recommended inspection. The AREVA format provides this information under a single heading. The Westinghouse format divides the methodology into four separate subheadings: "Goal," "Data Requirements," "Analysis," and "Acceptance Criteria." For the Westinghouse sections, unless otherwise indicated by specific terminology, "operating loads" refers to any loads generated under normal operating conditions.

WCAP- 17096-NP December 2009 Revision 2

4-2 4.5 RECOMMENDED APPROACH In general, implementation of the methodology and data requirement recommendations can require a combination of plant-specific and cooperative actions. Because the MRP-227 recommendations tend to be very design specific, generic actions that apply to the entire PWR fleet are not expected. However, it may be possible to define cooperative activities relevant to rational subgroups such as plants designed by a single vendor.

WCAP- 17096-NP December 2009 Revision 2

5-1 5 B&W PLANT DESIGN RESULTS 5.1 METHODOLOGIES A summary of the guidelines and methodology for determining acceptance criteria and the needed data requirements suggested by the expert panel for each of the Primary and Expansion component items identified in MRP-227 for the operating B&W-design reactor vessel internals is provided in Appendix A.

Appendix B provides logic charts illustrating the evaluation path and potential disposition options for relevant inspection conditions for each of the Primary and Expansion component items identified in MRP-227.

5.2 GENERIC ACCEPTANCE CRITERIA Each of the Primary and Expansion component items was evaluated to determine those for which analyses would be practicable to develop generic acceptance criteria. Differing unit loads, transients, materials, etc. were considered in identifying those component items that could be analyzed on a generic basis. The AREVA analysis indicates actions that might be used to define nondestructive examination (NDE) acceptance standards and actions that could support analytical evaluations. Table 5-1 below provides the results of this effort.

WCAP- 17096-NP December 2009 Revision 2

5-2 Table 5-1 Applicability of Potential Generic Acceptance Criteria for B&W-Design Primary and Expansion Component Items Develop Generic Acceptance Criteria?("

NDE Component Item Analytical Standard Comments Primary Items Plenum Cover Assembly & Core Yes Yes Support Shield Assembly Plenum cover weldment rib pads Plenum cover support flange CSS top flange Core Support Shield Assembly Yes Yes Analytical efforts applicable to both CSS cast outlet nozzles ONS-3 and DB could be performed, although unit-specific analytical efforts (Applicable to ONS-3 and DB may provide additional margin for one or only)(2) both units.

Core Support Shield Assembly Yes Yes A unit-specific bypass analytical effort is CSS vent valve discs 2

() required for DB, since the number of vent valves is different.

Core Support Shield Assembly Yes Yes Analytical efforts could be performed on a CSS vent valve top retaining ring generic basis for all B&W units, although unit-specific analytical efforts may provide CSS vent valve bottom retaining ring additional margin (particularly for DB).

CSS vent valve disc shaft or hinge pin Core Support Shield Assembly No Yes Due to the variations in bolt materials used and loadings among the units, unit-specific Upper core barrel (UCB) bolts and analytical efforts are required. The generic their locking devices efforts have already been completed in the PWROG PA-MSC-350 work.

Core Barrel Assembly No Yes Due to the variations in bolt materials used and loadings among the units, unit-specific Lower core barrel (LCB) bolts and analytical efforts are required. The generic their locking devices efforts have already been completed in the PWROG PA-MSC-350 work.

Core Barrel Assembly Yes N/A There are two designs (ONS-1 and TMI-1 are one design, and the other five units are Baffle-to-former bolts the second design).

Core Barrel Assembly Yes Yes Baffle plates WCAP- 17096-NP December 2009 Revision 2

5-3 Table 5-1 Applicability of Potential Generic Acceptance Criteria for B&W-Design Primary and (cont.) Expansion Component Items Develop Generic Acceptance Criteria?('

NDE Component Item Analytical Standard Comments Core Barrel Assembly N/A Yes There are two designs (ONS-1 and TMI-l Locking devices, including locking are second the one design and the other five units are design).

and welds, of baffle-to-former bolts internal baffle-to-baffle bolts Lower Grid Assembly Yes Yes Alloy X-750 dowel-to-guide block welds Incore Monitoring Instrumentation Yes Yes (VIMI) Guide Tube Assembly IMI guide tube spiders IMI guide tube spider-to-lower grid rib section welds Expansion Items Upper or Lower Grid Assembly Yes Yes Alloy X-750 dowel-to-upper grid fuel assembly support pad welds or Alloy X-750 dowel-to-lower grid fuel assembly support pad welds Control Rod Guide Tube Assembly No Yes Reactivity analyses are dependent upon fuel loading and must be performed on a CRGT spacer castings unit-specific basis.

Core Barrel Assembly Yes Yes Analytical efforts for the UTS bolt failures could be performed on a generic basis for Upper thermal shield (UTS) bolts and all units except TMI- 1, although use of their locking devices unit-specific loadings could reduce the conservatism for some units.

Core Barrel Assembly No No There are only two units: one has Surveillance specimen holder tube studs/nuts and the other has bolts.

(SSHT) studs/nuts (CR-3) or bolts (DB) and their locking devices Core Barrel Assembly Yes N/A Core barrel cylinder (including vertical and circumferential seam welds)

Former plates WCAP- 17096-NP December 2009 Revision 2

5-4 Table 5-1 Applicability of Potential Generic Acceptance Criteria for B&W-Design Primary and (cont.) Expansion Component Items Develop Generic Acceptance Criteria?")

NDE Component Item Analytical Standard Comments Core Barrel Assembly Yes N/A There are two designs (ONS- 1 and TMI- 1 are one design, and the other five units are Baffle-to-baffle bolts the second design).

Core barrel-to-former bolts Core Barrel Assembly Yes N/A Locking devices, including locking welds, for the external baffle-to-baffle bolts and core barrel-to-former bolts Lower Grid Assembly Yes Yes Lower grid fuel assembly support pad items: pad, pad-to-rib section welds, Alloy X-750 dowel, cap screw, and their locking welds (Note: The pads, dowels, and cap screws are included because of TE/IE of the welds.)

Lower Grid Assembly No No TMI unit-specific Lower grid shock pad bolts and their locking devices (TMI only)

Lower Grid Assembly No Yes Due to the variations in stud/nut or bolt Lower thermal shield studs/nuts oruntui-pcfcalyearrqied materials used and loadings among the bolts (LTS) and their locking devices units, unit-specific analyses are required.

Flow Distributor Assembly Yes Yes Analytical efforts for the FD bolts could be Flow distributor (FD) bolts and their performed on a generic basis (for all units Flowkingdirib r (except TMI-1), although unit-specific locking devices analyses could decrease the conservatism for some units.

Notes:

I. Analytical efforts include finite element analysis or fracture mechanics analysis. An NDE inspection standard contains examples of acceptable and unacceptable visual indications or ultrasonic testing flaw sizes.

2. These items may potentially be removed from examination if a records search identifies that actual material heats used for fabrication could be screened out as not being susceptible to the thermal aging degradation mechanism.

WCAP- 17096-NP December 2009 Revision 2

6-1 6 COMBUSTION ENGINEERING AND WESTINGHOUSE PLANT DESIGN RESULTS 6.1 METHODOLOGIES Datasheets outlining the guidelines and methodology for determining acceptance criteria and the needed data requirements suggested by the expert-panel for each of the Primary and Expansion component items in CE and Westinghouse plants are provided Appendices C and E. CE recommendations are contained in Appendix C. Westinghouse plant recommendations are contained in Appendix E.

Appendices D and F provide logic charts illustrating the evaluation path and potential disposition options for relevant inspection conditions for each of the Primary and Expansion component items identified in MRP-227.

6.2 GENERIC ACCEPTANCE CRITERIA Comments on the analysis approach for each component are included in the final section of the datasheets included in Appendices C and E. These recommendations, which include any actions that might be taken on a generic basis, are summarized in Tables 6-1 and 6-2.

WCAP- 17096-NP December 2009 Revision 2

6-2 Table 6-1 Applicability of Potential Generic Acceptance Criteria for CE-Design Primary and Expansion Component Items CE Component Approach CE-1D: 1 Core Shroud Bolts No generic effort required. Only two plants are affected.

CE-ID 1.1 Barrel Shroud Bolts No generic effort required. Only two plants are affected.

CE-ID: 1.2 Core Support Column Bolts Generic program to share first-of-a-kind effort. (See W-ID: 2.1)

• Pilot analysis of lower support structure to identify critical issues.

• Expect final acceptance based on plant-specific analysis.

CE-ID: 2 Core Shroud Plate-Former Plate Weld Expect calculation to be plant specific.

• Define general load conditions at weld seams.

• Define K-solution for loading at weld seams.

CE-ID: 2.1 Remaining Axial Welds Plant-specific analysis.

• Require flaw tolerance handbook/methodology based on flaw location and direction.

CE-ID: 3 Shroud Plates (Full Height) No generic analysis: Only one utility with this design.

CE-ID: 3.1 Remaining Axial Welds; Ribs and Rings No generic analysis: Only one utility with this design.

CE-ID: 4 Core Shroud Assembly (Bolted) FMEA should address plant-specific practices and priorities. Some generic work possible to outline issues and options to be addressed in FMEA.

CE-ID: 5 Core Shroud Assembly (Welded) Generic efforts to support inspection.

• Extension of MRP model to look at relationship between swelling and deformation at seam.

Guideline for issues to be addressed in plant-specific FMEA.

CE-ID: 6 Upper Core Support Barrel Flange Weld Plant-specific analysis.

• Similar to Ginna pilot plant experience. (See W-ID: 3)

CE-ID: 6.1 Lower Core Barrel Flange Plant-specific analysis.

• Require flaw tolerance handbook/methodology based on flaw location and direction.

MRP-2 10 may have limited relevance.

WCAP- 17096-NP December 2009 Revision 2

6-3 Table 6-1 Applicability of Potential Generic Acceptance Criteria for CE-Design Primary and (cont.) Expansion Component Items CE Component Approach CE-ID: 6.2 Remaining Core Barrel Assembly Welds Plant-specific analysis. (See item CE-ID: 6.1)

• Require flaw tolerance handbook/methodology based on flaw location and direction.

• MRP-2 10 may have limited relevance.

CE-ID: 7 Core Support Barrel Lower Flange Weld TLAA (plant specific)

Potential flaw analysis if inspection required.

• Require flaw tolerance handbook/methodology based on flaw location and direction.

• MRP-2 10 may have limited relevance.

CE-ID: 8 Core Support Plate TLAA (plant specific)

CE-ID: 9 Upper Fuel Alignment Plate TLAA (plant specific - applies to one utility)

CE-ID: 10 Instrument Guide Tubes Pass/Fail inspection with established minimum number of instrumentation tubes. Based directly on plant specifications.

CE-ID: 10.1 Remaining Instrument Guide Tubes Pass/Fail inspection with established minimum number of instrumentation tubes. Based directly on plant specifications. (See CE-ID: 10)

CE-ID: II Deep Beams TLAA (plant specific - applies to one utility)

WCAP- 17096-NP December 2009

. Revision 2

6-4 Table 6-2 Applicability of Potential Generic Acceptance Criteria for Westinghouse-Design Primary and Expansion Component Items Westinghouse Component Approach W-ID: 1 Control Rod Guide Tube (CRGT) Assembly Generic work ongoing under PWROG program Guide Plates (Cards)

• Validate and/or modify linear volumetric wear rate model.

Potential extension Alternative justification that allows wear through ligament in one or more cards.

W-ID: 2 CRGT Lower Flange Weld Plant-specific analysis due to large variety of sizes and designs. There may be some potential for smaller plant groupings.

W-ID: 2.1 Lower Support Column Bodies (Cast) Generic program to share first-of-a-kind effort.

• Pilot analysis of lower support structure to identify critical issues.

• Expect final acceptance based on plant-specific analysis.

W-ID: 2.2 Bottom-mounted Instrumentation Colunm Pass/Fail inspection with established minimum number Bodies of instrumentation tubes. Based directly on plant specifications.

W-ID: 3 Upper Core Barrel Flange Weld Plant-specific analysis.

• Ginna provides pilot plant experience in the creation of generic acceptance criteria.

• May be able to group plants by design.

W-ID: 3.1 Other Core Barrel Welds Plant-specific analysis.

• Require flaw tolerance handbook/methodology based on flaw location and direction.

MRP-2 10 may have limited relevance.

W-ID: 3.2 Lower Support Column Bodies (Non-cast) Generic program to share first-of-a-kind effort. (See W-ID: 2.1)

• Pilot analysis of lower support structure to identify critical issues.

• Expect final acceptance based on plant-specific analysis.

W-ID: 4 Baffle-edge Bolts FMEA should address plant-specific practices and priorities. Some generic work possible to outline issues and options to be addressed in FMEA.

W-ID: 5 Baffle-former Bolts Generic work completed in previous PWROG program.

W-ID: 5.1 Barrel-former Bolts Generic work completed in previous PWROG program.

WCAP- 17096-NP December 2009 Revision 2

6-5 Table 6-2 Applicability of Potential Generic Acceptance Criteria for Westinghouse-Design Primary (cont.) and Expansion Component Items Westinghouse Component Approach W-ID: 5.2 Lower Support Column Bolts Generic program to share first-of-a-kind effort. (See W-ID: 2-1)

• Pilot analysis of lower support structure to identify critical issues.

• Expect final acceptance based on plant-specific analysis.

W-ID: 6 Baffle-former Assembly FMEA should address plant-specific practices and priorities. Some generic work possible to outline issues and options to be addressed in FMEA.

W-ID: 7 Internals Hold-down Spring Value determined by plant-specific design requirements.

W-ID: 8 Thermal Shield Flexures None: Plant-specific analysis.

WCAP- 17096-NP December 2009 Revision 2

7-1 7 REFERENCES

1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0). EPRI, Palo Alto, CA: 2008. 1016596.

2. Materials Reliability Program: Screening, Categorization, and Ranking of B&W-Designed PWR Internals Component Items (MRP- 189 Rev. 1). EPRI, Palo Alto, CA: 2009. 1018292.

3. Materials Reliability Program: Failure Modes, Effects, and Criticality Analysis of B&W-Designed PWR Internals (MRP-190). EPRI, Palo Alto, CA: 2006. 1013233.

4. Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Designs (MRP-191). EPRI, PaloAlto, CA: 2006. 1013234.

5. Materials Reliability Program: Functionality Analysis for B&W Representative PWR Internals (MRP-229 Rev. 1). EPRI, Palo Alto, CA: 2009. 1019090.

6. Materials Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (MRP-230). EPRI, Palo Alto, CA: 2008. 1016597.

7. Materials Reliability Program: Aging Management Strategies for B&W PWR Internals (MRP-231 Rev. 1). EPRI, Palo Alto, CA: 2009. 1019092.

8. Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232). EPRI, Palo Alto, CA: 2008. 1016593.

9. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Components," American Society of Mechanical Engineers, New York, NY: 2001 Edition, Plus 2003 Addenda, or later.

10. Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228). EPRI, Palo Alto, CA: 2009. 1016609.

11. Materials Reliability Program: Pressurized Water Reactor Issue Management Table, PWR-IMT Consequence of Failure (MRP-156). EPRI, Palo Alto, CA: 2005. 1012110.

12. U.S. Nuclear Regulatory Commission Report, NUREG-1801, Vol. 1, Rev. 1, "Generic Aging Lessons Learned (GALL) Report," September 2005.

13. Westinghouse Report, WCAP-1 5030-NP-A, Rev. 0, "Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions under Faulted Load Conditions,"

March 2, 1999.

WCAP- 17096-NP December 2009 Revision 2

A-1 APPENDIX A B&W DESIGN PRIMARY AND EXPANSION COMPONENT ITEM ACCEPTANCE CRITERIA METHODOLOGY AND DATA REQUIREMENTS A.1 PRIMARY COMPONENT ITEMS Acceptance criteria methodology and data requirements for each of the Primary component items are summarized in this appendix. A separate sub-section is provided for each component item using the following format:

• Primary component item information extracted directly from Table 4-1 of MRP-227.

• This information is in tabular form and contains the item name, unit applicability, failure effect, failure mechanism(s), expansion link(s), examination method, examination frequency, and examination coverage.

• Component item function(s), including whether or not it has a core support safety function.

• Observable effect(s).

• Methodology for development of acceptance criteria.

• Data requirements for development of acceptance criteria.

• Existing documents (e.g., PWROG or AREVA).

Note that repairs or replacements are potential options for various components that are not summarized in this appendix, but are included in Appendix B.

WCAP- 17096-NP December 2009 Revision 2 I I I

A-2 Core Clamping Items

• Effect Expansion, , Examination 'ý,Examination Item Applicability (Mechanism) Link Method/Frequency, Coverage' Plenum Cover All plants Loss of None. One-time physical Determination of Assembly & material and measurement no differential height Core Support associated loss later than two of top of plenum Shield Assembly of core refueling outages rib pads to reactor clamping from the beginning vessel seating Plenum cover pre-load of the license surface, with weldment rib pads (Wear) renewal period. plenum in reactor Plenum cover Perform subsequent vessel.

support flange visual (VT-3) See MRP-231 CSS top flange examination on the Figure 3-1.

10-year ISI interval.

The potential exists for this to be a high wear location with subsequent reduction or loss of core clamping pre-load. This is a unique configuration with the B&W units and no known OE history indicating wear currently exists.

Component Item Function The purpose of the core clamping load is to stabilize and significantly restrict the rigid body pendulum motion of the core support and plenum assemblies. In other words, the clamping action prevents rigid body rotation at the interface area. The clamping action does not have a direct core support safety function. Loss of clamping would undoubtedly lead to core barrel motions that would eventually lead to a reactor shut down.

Observable Effects:

A one-time physical measurement is to be obtained. The interference measurement to the nearest 0.001 inch is to be recorded at eight locations at approximately 45-degree intervals. Subsequent follow-up visual (VT-3) examinations are to be obtained during the ASME Code B-N-3 10-year ISI activities.

ONS has performed physical measurements on a unit-specific basis and no measurable wear has been observed. TMI-1 plans to obtain the data during the Fall 2009 outage. The remaining units have not completed the measurement to date.

The physical measurement is performed to determine the differential height from the top of the plenum cover assembly weldment rib pads to the reactor vessel seating surface. The measurement is the stack-up of the core support flange and the plenum support area versus the reactor vessel support ledge. The measurement must be taken without fuel in the reactor to eliminate the effect of the fuel hold-down springs. This interference fit was measured during original site assembly and also as-built measurements were taken of the piece parts after fabrication. The interference fit ranged between 0.008 and zero for the operating units.

WCAP- 17096-NP December 2009 Revision 2

A-3 Identification of an unacceptable condition is the precursor to perform follow-on investigations or analysis to establish the extent of degradation. The subsequent VT-3 inspections are to look for wear on the top and bottom surface of the CSS top flange, bottom surface of the plenum cover assembly support flange, and the top surface of the plenum cover assembly weldment rib pads.

Possible Examination Outcomes:

• No wear observed (data falls within scatter of original measurements)

• Wear of some extent is observed (one location, several locations, etc.)

• Other relevant conditions identified (e.g., missing rib pad)

Methodology and Data Requirements:

The acceptance criterion is based on engineering judgment, and is defined as a reduction of no greater than 0.004 inch compared to the original as-built data. The criterion of 0.004 inch reduction in interference is not be construed as an indication of inadequate clamping, but an indication that the surface conditions have changed since the unit was put into service. Additional inspection, using VT-3 for example, will be required to verify that wear is actually occurring.

The general analytical methodology to be used for determining the wear acceptance criterion involves the following steps and inputs:

Determine the minimum core clamping preload required

- This requires the differential pressure distribution on the core support cylinder due to reactor coolant flow

- This also includes evaluation for different minimum pre-loads versus time at different operating conditions Determine an uncertainty margin, which includes both input pressure differential and measurement error Determine how the clamp load varies with operating conditions Develop a wear estimate for time from discovery of wear to time for required remediation The general methodology to be used for VT-3 acceptance criteria for these component items will be development of an NDE inspection standard that contains examples of acceptable and unacceptable visual indications and mockups for the VT-3 inspection of wear. Input information needed includes:

Identification of the most. likely signs and locations of wear that can be inspected Identification of what visual examination wear indications are considered rejectable and would require additional dimensional examination and evaluation WCAP- 17096-NP December 2009 Revision 2

A-4 Identification of any additional examination results that are anticipated

- Identify general acceptance criteria for the additional items expected to be in the VT-3 examination field of vision Analytical efforts could be performed on a generic basis. The NDE inspection standard could also be developed generically.

Existing Documentation:

In the design stage, a design clamping load at operational conditions was established based on the flow uplift loading and horizontal forces developed by various pump combinations. An additional study in 1974 indicated that the clamping value was marginal using the reactor vessel stud pre-load identified in the reactor vessel instruction manual. A new pre-load was incorporated into the design, which provided satisfaction of the criteria at operating conditions but indicated there may be some loss of clamping at low temperatures (approximately < 300'F). It was verified by observation that the potential loss of pre-load was not leading to short-term degradation of the clamping surfaces.

No acceptable evaluation or analysis has been completed to date for determining a re-inspection schedule.

- A wear rate estimate would be needed to project the wear over one cycle or more for continued operation.

What observations trigger examination into the Expansion category?

There are no expansion component item links for this examination.

WCAP- 17096-NP December 2009 Revision 2

A-5 Core Support Shield Cast Outlet Nozzles Effect --Expansion Examination Examination itItem ApplicabiH' (Mechanism) Link Method/Frequency .Coverage Core Support ONS-3, DB Cracking CRGT spacer Visual (VT-3) 100% of Shield (TE), castings examination during accessible Assembly including the the next 10-year ISI. surfaces.

CSSdetection of Subsequent See MRP-231 outlet nozzles irface examinations on the Figure 3-9.

irregularities, 10-year ISI interval.

such as damaged or fractured material CSS cast outlet nozzles are subject to thermal aging embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the materials degraded fracture toughness, such a condition could potentially lead to cracking. There is no known history of OE identifying cracking of CASS material in PWR reactor vessel internals applications.

These items may potentially be removed from examination if a records search identifies that actual material heats used for fabrication could be screened out as not being susceptible to the thermal aging degradation mechanism.

Component Item Function Degradation of the outlet nozzles could result in a core cooling issue under normal operation because of increased core bypass flow and a reduction in margin to DNB (see MRP-157). The outlet nozzles do not have a core support safety function; however, they do have a safety function to control bypass around the core during a loss-of-coolant-accident (LOCA).

Observable Effects:

A visual (VT-3) examination of the outlet nozzles is to be performed. Subsequent visual examinations are to be performed during the ASME Code B-N-3 10-year ISI activities.

The outlet nozzles are being examined to detect damage either in the form of a precursor to the loss of material or a piece or section of the material that has fractured and is currently missing. The location that potentially contains the highest tensile stresses is near the heat-affected-zone (HAZ) of the outlet nozzle weld-to-core support shield cylinder.

Possible Examination Outcomes:

0 No relevant conditions identified S One or more areas are identified with crack-like indications 0 One or more areas are identified with missing material WCAP- 17096-NP December 2009 Revision 2

A-6 Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the CSS cast outlet nozzles involves the following steps and inputs:

• Perform a bypass analysis to justify that sufficient DNB exists in the degraded condition

• A VT-1, ET, or UT examination may be needed to determine if flaws are emanating from the location where missing material may be identified The general methodology to be used for acceptance criteria for these component items will be development of an NDE inspection standard that contains examples of acceptable and unacceptable visual indications and mockups for the VT-3 inspection of cracking. Input information needed includes:

• Identification of the most likely locations of surface irregularities, such as damaged or fractured material

• Identification of what visual examination indications are considered rejectable and would require additional examination and evaluation Analytical efforts applicable to both ONS-3 and DB could be performed, although unit-specific analytical efforts may provide additional margin for one or both units. The NDE inspection standard could also be developed generically.

Existing Documentation:

• CLB loadings (normal and faulted condition) are available, but a records search may need to be performed to identify them

• No acceptable evaluation or analysis has been completed to date for determining a re-inspection schedule What observations trigger examination into the Expansion category?

• Confirmed evidence of relevant conditions for a single CSS cast outlet nozzle shall require expansion to the CRGT spacer castings by the completion of the next refueling outage December 2009 WCAP- 7096-NP WCAY- 117096-NP December 2009 Revision 2

A-7 Core Support Shield Vent Valve Discs Effect Expansion' Examination Examination Item Applicability .(Mechanism) . Link Method/Freque ncy Coverage Core Support All plants Cracking (TE), CRGT spacer Visual (VT-3) 100% of accessible Shield including the castings examination during surfaces.

Assembly detection of the next 10-year ISI.

surface (See BAW-2248A, CSS vent valve irregularities, Subsequent page 4-3 and discs such as examinations on the Table 4-1.)

(Note 1) damaged or 10-year ISI interval. See MRP-231 fractured Figures 3-10 and material 3-11.

Notes:

1. A verification of the operation of each vent valve shall also be perforned through manual actuation of the valve. Verify that the valves are not stuck in the open position and that no abnormal degradation has occurred. Examine the valves for evidence of scratches, pitting, embedded particles, variation in coloration of the seating surfaces, cracking of lock welds and locking cups, jack screws for proper position, and wear. The frequency is defined in each unit's technical specifications or in their pump and valve in-service test programs (see AREVA doc. BAW-2248A, page 4-3, and Table 4-1).

Vent valve discs are subject to thermal aging embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the materials degraded fracture toughness, such a condition could potentially lead to cracking. There is no known history of OE identifying cracking of CASS material in PWR reactor vessel internals applications.

These items may potentially be removed from examination if a records search identifies that actual material heats used for fabrication could be screened out as not being susceptible to the thermal aging degradation mechanism.

Comnonent Item Function Vent valves are passive devices that have no function during normal operation. The vent valve discs do not have a core support safety function; however, they do have a safety function in that degradation of the vent valve discs, which would prevent the vent valve from opening, could result in loss of the vent valve function during a large break loss-of-coolant-accident (LOCA).

Observable Effects:

A visual (VT-3) examination of the vent valve discs is to be performed. Subsequent visual examinations are to be performed during the ASME Code B-N-3 10-year ISI activities.

The vent valve discs are being examined to detect damage either in the form of a precursor to the loss of material, or, a piece or section of the material that has fractured or is currently missing.

Possible Examination Outcomes:

No relevant conditions identified WCAP- 17096-NP December 2009 Revision 2

A-8

• One or more areas are identified with crack-like indications

• One or more areas are identified with missing material Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the CSS vent valve discs involves the following steps and inputs:

• Perform a bypass analysis to justify that sufficient DNB exists in the degraded condition

• Perform an analysis to show that failure of the vent valve disc will not result in loss of function

• A VT-1, ET, or UT examination may be needed to determine if flaws are emanating from the location where missing material may be identified The general methodology to be used for acceptance criteria for these component items will be development of an NDE inspection standard that contains examples of acceptable and unacceptable visual indications and mockups for the VT-3 inspection of cracking. Input information needed includes:

• Identification of the most likely locations of surface irregularities, such as damaged or fractured material

• Identification of what visual examination indications are considered rejectable and would require additional examination and evaluation Analytical efforts could be performed on a generic basis for all B&W-design units; however, the number of vent valves for DB is different from the rest of the units, which would require a unit-specific bypass analytical effort. The NDE inspection standard could also be developed generically.

Existing Documentation:

• CLB loadings (normal and faulted condition) are available, but a records search may need to be performed to identify them What observations trigger examination into the Expansion category?

• Confirmed evidence of relevant conditions (damage, grossly cracked, or fractured material) in two or more vent valve discs shall require expansion to the CRGT spacer castings by the completion of the next refueling outage WCAP- 17096-NP December 2009 Revision 2

A-9 Core Support Shield Vent Valve Retaining Rings and Disc Shaft Effect Expansion Examination Examination Item Applicability (Mechanism) Link Method/Frequency C'v'rage:

Core Support All plants Cracking None Visual (VT-3) 100% of accessible Shield Assembly (TE), examination during surfaces.

CSS vent CS valve top etvletpdetection including of(SeAW28, the the next 10-year ISI. (See BAW-2248A, retaining ring surface Subsequent page 4-3 and examinations on the Table 4-1.)

CSS vent valve irregularities, 10-year 151 interval.

bottom retaining such as See MRP-231 ring damaged, Figures 3-10 and fractured 3-11.

CSS vent valve material, or disc shaft or hinge missing items pin (Note 1)

Notes:

1. A verification of the operation of each vent valve shall also be performed through manual actuation of the valve. Verify that the valves are not stuck in the open position and that no abnormal degradation has occurred. Examine the valves for evidence of scratches, pitting, embedded particles, variation in coloration of the seating surfaces, cracking of lock welds and locking cups, jack screws for proper position, and wear. The frequency is defined in each unit's technical specifications or in their pump and valve in-service test programs (see AREVA doc. BAW-2248A, page 4-3, and Table 4-1).

Vent valve top and bottom retaining rings and the disc shaft (or, hinge pin) are subject to thermal aging embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the materials degraded fracture toughness, such a condition could potentially lead to cracking. Although there have been instances of failures of precipitation-hardenable materials in other applications, there is no known history of OE identifying cracking in PWR reactor vessel internals applications.

Component Item Function Vent valves are passive devices that have no function during normal operation. The vent valve top and bottom retaining rings and the disc shaft (or, hinge pin) do not have a core support safety function; however, they do have a safety function in that degradation of the vent valve top and bottom retaining rings and the disc shaft (or, hinge pin), which would prevent the vent valve from opening, could result in loss of the vent valve function during a large break loss-of-coolant-accident (LOCA).

Observable Effects:

A visual (VT-3) examination of the accessible surfaces of the vent valve top and bottom retaining rings and the disc shaft (or, hinge pin) is to be performed. Subsequent visual examinations are to be performed during the ASME Code B-N-3 10-year ISI activities.

The vent valve top and bottom retaining rings and the disc shaft (or, hinge pin) are being examined to detect damage either in the form of a precursor to the loss of material, or, a piece or section of the material WCAP- 17096-NP December 2009 Revision 2

A-10 that has fractured or is currently missing, particularly, in the areas where high stresses exist. For example, with the retaining rings, at the locations where the jacking screws are connected.

Possible Examination Outcomes:

• No relevant conditions identified

• Observation that either retaining ring or disc shaft (or, hinge pin) is not in the correct position

• One or more areas are identified with crack-like indications

• One or more areas are identified with missing material Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the vent valve top and bottom retaining rings and the disc shaft (or, hinge pin) involves the following steps and inputs:

• Perform analysis to show that failure of vent valve items will not result in loss of function

• Perform a bypass analysis to justify that sufficient DNB exists in the degraded condition

• A VT- 1, ET, or UT examination may be needed to determine if flaws are emanating from the location where missing material may be identified The general methodology to be used for acceptance criteria for these component items will be development of an NDE inspection standard that contains examples of acceptable and unacceptable visual indications and mockups for the VT-3 inspection of cracking. Input information needed includes:

• Identification of the most likely locations of surface irregularities, such as damaged, fractured material, or missing items

• Identification of what visual examination indications are considered rejectable and would require additional examination and evaluation Analytical efforts could be performed on a generic basis for all B&W units, although unit-specific analytical efforts may provide additional margin (particularly for DB). The NDE inspection standard could also be developed generically.

Existing Documentation:

• CLB loadings (normal and faulted condition) are available, but a records search may need to be performed to identify them

• Manufacturing and material data need to be identified to determine chemical composition and an assessment of the actual susceptibility to thermal aging embrittlement

• No acceptable evaluation or analysis has been completed to date for determining a re-inspection schedule WCAP- 17096-NP December 2009 Revision 2

A-1I What observations trigger examination into the Expansion category?

There are no expansion items for these component items.

WCAP- 17096-NP December 2009 Revision 2

A-12 Upper Core Barrel Bolts and Locking Devices

• Effect,, Expansion . Examination Examination, Item Applicability (Mechanism) Link Method/Frequency Coverage.>

Core Support All plants Cracking (SCC) LCB (Note 1) Volumetric 100% of Shield examination (UT) of accessible Assembly UTS, LTS, and FD bolts* the two bolts within refueling bolts.

Upper core SSo blsues See MRP-231 barrel (UCB) SSHT bolts outages from Fige 3-7.

(CR-3 and DB 1/1/2006 or next bolts and their locking only) 10-year ISI interval, whichever is first.

devicesLower grid shock pad bolts Subsequent (TMI- 1 only) examination to be determined after evaluating the baseline results.

Visual (VT-3) examination of bolt locking devices on the 10-year ISI interval.

Note:

1. Expansion to LCB applies if the required Primary examination of LCB has not been performed as scheduled in this table.

There is a potential for intergranular stress corrosion cracking (IGSCC) of Alloy A-286 and Alloy X-750 bolting. Past B&W failure history exists with the original Alloy A-286 bolt materials in B&W-design units and with applications of Alloy X-750 material within the nuclear industry (in general). Currently, there are no known failures with any of the replacement bolts (Alloy A-286 or Alloy X-750) in the operating B&W-design units or with the original Alloy X-750 (installed at TMI-1 only) in service in the operating B&W-design units.

Component Item Function The UCB joint carries the entire weight of the core and the majority of the weight of the reactor vessel internals. The upper core barrel bolts have a core support safety function in that should the joint fail, the core and internals could drop, coming to rest on the guide lugs welded to the inside wall of the reactor vessel.

Observable Effects:

A volumetric examination (UT) of the bolts and a visual (VT-3) examination of the bolt locking devices is to be performed.

Mockups and qualification efforts exist for UCB and LCB bolts from the PWROG work (PA-MSC-350) and additional Duke Energy efforts in 2007-2008.

WCAP- 17096-NP December 2009 Revision 2

A-13 Cracking of the bolts is the main concern and the locking devices are to be examined to identify if any are distorted, loose, broken, or missing.

The PWROG work (PA-MSC-350) also evaluated the potential information that could be determined from only a visual examination of the bolt and locking devices (see AREVA document 51-9081184-001).

Possible Examination Outcomes:

Cracking is anticipated to occur at the head-to-shank area where the peak tensile stress exists (i.e., a SCF exists) and OE has shown them to crack at this location in the past, although it may also occur in the shank thread region where high tensile stress is possible too.

• No relevant conditions identified

• Relevant conditions (i.e., crack-like indications, either completely cracked or partially cracked; or non-interpretable UT indications, such as no back wall reflection or multiple reflections with no crack-like indication that is most likely caused by a large or duplex grain size) are identified

- *One or a few bolts (exact number is unit-specific) are identified with relevant indications

- More than a few bolts (exact number is unit-specific) are identified with relevant indications Locking Devices:

• No relevant conditions identified

• One or two are identified with damage or are missing

• More than two bolt locking devices are identified with damage or are missing Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the bolts involves the following steps and inputs if relevant conditions have been identified in the UCB bolts:

• A finite element model (FEM) is to be developed for the local geometry with contact conditions, pretension elements, loads and boundary conditions

• A thermal analysis is to be performed

- Determines bolt temperatures and temperature gradients for normal operating conditions

• A structural analysis is to be performed in which failed bolts are inactive

- Stress concentration factors are calculated to determine the peak stresses at the bolt head-to-shank fillet region under normal operating conditions WCAP- 17096-NP December 2009 Revision 2

A-14 Analysis is performed for all loads and load combinations required for an ASME evaluation (stress limits for threaded structural fasteners in subsection NG and Appendix F)

The effect of failed or missing bolts on overall effective core barrel stiffness is evaluated A change of no more than 20% in stiffness when subjected to LOCA loads is acceptable (within the limits of other uncertainties accounted for in the evaluation of LOCA loadings)

This corresponds to approximately 10% change in fundamental frequency of the structure

- An evaluation of joint stability (or, openness) is also to be performed Representative rejected UCB bolts are to be removed for laboratory testing to confirm the UT inspection results and IGSCC mechanism, or to identify any other failure mechanism(s)

NOTE: One alternative to removing representative UT rejected bolts for laboratory examination is to re-inspect all bolts at the next refueling outage if continued operation for one additional fuel cycle can be supported by a technical evaluation. Other possible options such as replacement of a predefined bolt pattern may also be pursued.

Based on the results of UCB bolt UT inspection and laboratory test results, perform an evaluation to assess future UCB bolt failure potential. The changes to the peak stress at the bolt head-to-shank fillet region as a result of the identified failures should be included for evaluation of increased susceptibility to SCC.

Incorporate the effect of future UCB bolt failure into the operability evaluation and re-inspection requirement The general methodology to be used for acceptance criteria for the locking devices will be development of an inspection standard that contains examples of acceptable and unacceptable locking device visual indications. The acceptance of locking devices is evaluated in two ways: a) observations with "failed" bolts and b) observations with all bolts intact. Observations of damaged locking devices with all bolts intact represent a condition very different from that of locking device damage at a bolt location that is failed. In addition, a damaged or missing welded locking clip versus a crimpled locking cup potentially represents different initiating phenomena that need to be evaluated.

Due to the variations in bolt materials used and loadings among the units, unit-specific analytical efforts are required. [The generic efforts have already been completed in the PWROG PA-MSC-350 work.] The NDE inspection standard could be developed generically.

WCAP- 17096-NP December 2009 Revision 2

A-15 Existing Documentation:

• A generic UCB bolt stress analysis was completed in the MRP reactor internals project for the three ONS units, CR-3, and ANO-1, which was subsequently used in the PA-MSC-0350 work noted below (see AREVA document 32-9095906).

FEM and thermal analysis have been developed by the PWROG project (PA-MSC-0350, see AREVA document 51-9089393-000) to evaluate failures for use in evaluating an acceptable failure pattern or number of failed UCB and LCB bolts allowed for continued operation, and for use in preparing a JCO.

One or a few bolts could be identified as failed (non-interpretable bolt UT signal; inaccessible bolt for UT; locking device observed to be missing, non-functional, or removed; partially cracked bolt; or completely cracked bolt) and shown to be acceptable with no further action needed A number of bolts (TBD) identified as failed (non-interpretable UT signal, partially cracked, or completely cracked) would initiate replacement activities and lead into the expansion category (see AREVA document 51-9087042-000 for initial methodology developed by PWROG project)

Some unit-specific analyses for several assumed inspection cases have been performed for each unit in the PWROG project (PA-MSC-350), but unit-specific analyses may need to be re-performed for the actual inspection result (see AREVA document 51-9089393-000).

If inspection results indicate no relevant indications of failure and calculated peak stresses are below the bolt material yield strength, SCC is not expected to initiate and an inspection during the next ASME Code B-N-3 I 0-year ISI interval is judged adequate If a relevant inspection condition is detected and confirmed by laboratory testing, a future bolt failure rate is needed for operability assessment and developing a future inspection frequency.

No acceptable evaluation or analysis methodology has been developed to date for assessing future bolt failures or determining a re-inspection schedule based on the inspection results.

A suggested approach, based on a combinatoric risk analysis, has been provided to the PWROG (see PWROG MSC December 2008 meeting minutes) that is based on unit-specific inspection results and unit-specific bolt structural analysis.

An NRC-accepted crack growth rate for Alloy A-286 or Alloy X-750 material is not currently available.

However, the PWROG project (PA-MSC-350, AREVA document 51-9079485-000) has identified some CGR data that is currently available. A CGR based life analysis has not been performed for any structural bolts. Based on the available CGR data, a life assessment based entirely on the CGR without considering crack initiation is judged unlikely to yield acceptable results.

WCAP- 17096-NP December 2009 Revision 2

A-16 What observations trigger examination into the Expansion category?

Cracking observed in 10% (12) of the UCB bolts (LCB bolts are only considered in the expansion if they have not already been inspected). Damage to locking devices for failed bolt locations is not unusual and would be anticipated; however, if damage to locking devices is observed in non-failed bolt locations, the second trigger criterion would also be used.

Observation of more than two locking devices damaged or missing (if no bolts are observed to be failed), pending additional evaluation as to the potential cause.

Should it trigger expansion to all remaining bolt rings or a tiered approach?

When an inspection triggers into the expansion, there is a unit-specific need to evaluate the results against the differences in materials used for the different locations and results from other unit inspections. For example, if failures are noted in the Alloy A-286 UCB bolts, but the UTS bolts are made from Alloy X-750 and no failures of this bolting material has been observed at any of the operating B&W units, a justification not to expand to this location should be possible. In addition, should failures be noted in a heat of Alloy A-286 used at one unit, expansion into the other units may need to be considered. However, one or more of the bolts with indications need to be removed for laboratory testing to confirm the IGSCC failure mechanism and stress analyses need to be performed for each of the expansion bolt locations.

Thus, expansion is not to be considered carte-blanche without additional evaluation.

WCAP- 17096-NP December 2009 Revision 2

A-17 Lower Core Barrel Bolts and Locking Devices Effect, Expansion, Examinati6n Examination Apliability (Meanism Link Method/Frequency Coverage' Core Barrel All plants Cracking UTS, LTS, and Volumetric 100% of accessible.

Assembly (SCC) FD bolts examination (UT) of bolts.

the bolts during the See MRP-231 Lower core SSHT bolts next 10-year ISI barrel (LCB) (CR-3 and DB interval from Figure 3-8.

bolts and only) 1/1/2006.

their locking devices Lower grid Subsequent shock pad bolts examination to be (TMI- 1 only) determined after evaluating the baseline results.

Visual (VT-3) examination of bolt locking devices on the 10-year ISI interval.

There is a potential for intergranular stress corrosion cracking (IGSCC) of Alloy A-286 and Alloy X-750 bolting. Past B&W failure history exists with the original Alloy A-286 bolt materials in B&W-design units and with applications of Alloy X-750 material within the nuclear industry (in general). Currently, there are no known failures with any of the replacement bolts (Alloy A-286 or Alloy X-750) in the operating B&W-design units or with the original Alloy X-750 (installed at TMI- 1 only) in service in the operating B&W-design units.

Component Item Function The LCB joint carries the entire weight of the core (but not the weight of the core barrel) and the weight of the lower reactor vessel internals. The lower core barrel bolts have a core support safety function in that should the joint fail, the core and lower internals could drop, coming to rest on the guide lugs welded to the inside wall of the reactor vessel.

Observable Effects:

A volumetric examination (UT) of the bolts and a visual (VT-3) examination of the bolt locking devices Mockups and qualification efforts exist for UCB and LCB bolts from the PWROG work (PA-MSC-350) and additional Duke Energy efforts in 2007-2008.

Cracking of the bolts is the main concern and the locking devices are to be examined to identify if any are distorted, loose, broken, or missing.

The PWROG work (PA-MSC-350) also evaluated the potential information that could be determined from only a visual examination of the bolt and locking devices (see AREVA document 51-9081184-001).

WCAP- 17096-NP December 2009 Revision 2

A-18 Possible Examination Outcomes:

Bolts Cracking is anticipated to occur at the head-to-shank area where the peak tensile stress exists (i.e., a SCF exists) and OE has shown them to crack at this location in the past, although it may also occur in the shank thread region where high tensile stress is possible too.

• No relevant conditions identified

• Relevant conditions (i.e., crack-like indications, either completely cracked or partially cracked; or non-interpretable UT indications, such as no back wall reflection or multiple reflections with no crack-like indication that is most likely caused by a large or duplex grain size) are identified

- One or a few bolts (exact number is unit-specific) are identified with relevant indications

- More than a few bolts (exact number is unit-specific) are identified with relevant indications Locking Devices

• No relevant conditions identified

• One or two are identified with damage or are missing

• More than two bolt locking devices are identified with damage or are missing Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the bolts involves the following steps and inputs if relevant conditions have been identified in the LCB bolts:

• A finite element model (FEM) is to be developed for the local geometry with contact conditions, pretension elements, loads and boundary conditions

• A thermal analysis is to be performed

- Determines bolt temperatures and temperature gradients for normal operating conditions

• A structural analysis is to be performed in which failed bolts are inactive

- Stress concentration factors are calculated to determine the peak stresses at the bolt head-to-shank fillet region under normal operating conditions

- Analysis is performed for all loads and load combinations required for an ASME evaluation (stress limits for threaded structural fasteners in subsection NG and Appendix F)

- An evaluation of joint stability (or, openness) is also to be performed WCAP- 17096-NP December 2009 Revision 2

A-19 Representative rejected LCB bolts are to be removed for laboratory testing to confirm the UT inspection results and IGSCC mechanism, or to identify any other failure mechanism(s)

NOTE: One alternative to removing representative UT rejected bolts for laboratory examination is to re-inspect all bolts at the next refueling outage if continued operation for one additional fuel cycle can be supported by a technical evaluation. Other possible options such as replacement of a predefined bolt pattern may also be pursued.

• Based on the results of LCB bolt UT inspection and laboratory test results, perform an evaluation to assess future LCB bolt failure potential. The changes to the peak stress at the bolt head-to-shank fillet region as a result of the identified failures should be included for evaluation of increased susceptibility to SCC.

Incorporate the effect of future LCB bolt failure into the operability evaluation and re-inspection requirement The general methodology to be used for acceptance criteria for the locking devices will be development of an inspection standard that contains examples of acceptable and unacceptable locking device visual indications. The acceptance of locking devices is evaluated in two ways: a) observations with "failed" bolts and b) observations with all bolts intact. Observations of damaged locking devices with all bolts intact represent a condition very different from that of locking device damage at a bolt location that is failed. In addition, a damaged or missing welded locking clip versus a crimpled locking cup potentially represents different initiating phenomena that need to be evaluated.

Due to the variations in bolt materials used and loadings among the units, unit-specific analytical efforts are required. [The generic efforts have already been completed in the PWROG PA-MSC-350 work.] The NDE inspection standard could be developed generically.

Existing Documentation:

A generic LCB bolt stress analysis was completed in the MRP reactor internals project for the three ONS units and ANO-1, which was subsequently used in the PA-MSC-0350 work noted below (see AREVA document 32-9095906).

FEM and thermal analysis have been developed by the PWROG project (PA-MSC-0350, see AREVA document 51-9089393-000) to evaluate failures for use in evaluating an acceptable failure pattern or number of failed UCB and LCB bolts allowed for continued operation, and for use in preparing a JCO.

- One or a few bolts could be identified as failed (non-interpretable bolt UT signal; inaccessible bolt for UT; locking device observed to be missing, non-functional, or removed; partially cracked bolt; or completely cracked bolt) and shown to be acceptable with no further action needed

- A number of bolts (TBD) identified as failed (non-interpretable UT signal, partially cracked, or completely cracked) would initiate replacement activities and lead into the WCAP- 17096-NP December 2009 Revision 2

A-20 expansion category (see AREVA document 51-9087042-000 for initial methodology developed by PWROG project)

Some unit-specific analyses for several assumed inspection cases have been performed for each unit in the PWROG project (PA-MSC-350), but unit-specific analyses may need to be re-performed for the actual inspection result (see AREVA document 51-9089393-000).

If inspection results indicate no relevant indications of failure and calculated peak stresses are below the bolt material yield strength, SCC is not expected to initiate and an inspection during the next ASME Code B-N-3 10-year ISI interval is judged adequate If a relevant inspection condition is detected and confirmed by laboratory testing, a future bolt failure rate is needed for operability assessment and developing a future inspection .frequency.

No acceptable evaluation or analysis methodology has been developed to date for assessing future bolt failures or determining a re-inspection schedule based on the inspection results.

- A suggested approach, based on a combinatoric risk analysis, has been provided to the PWROG (see PWROG MSC December 2008 meeting minutes) that is based on unit-specific inspection results and unit-specific bolt structural analysis.

An NRC-accepted crack growth rate for Alloy A-286 or Alloy X-750 material is not currently available.

However, the PWROG project (PA-MSC-350, AREVA document 51-9079485-000) has identified some CGR data that is currently available. A CGR based life analysis has not been performed for any structural bolts. Based on the available CGR data, a life assessment based entirely on the CGR without considering crack initiation is judged unlikely to yield acceptable results.

What observations trigger examination into the Expansion category?

Cracking observed in 10% (11) of the LCB bolts. Damage to locking devices for failed bolt locations is not unusual and would be anticipated; however, if damage to locking devices is observed in non-failed bolt locations, the second trigger criterion would also be used.

Observation of more than two locking devices damaged or missing (if no bolts are observed to be failed), pending additional evaluation as to the potential cause.

Should it trigger expansion to all remaining bolt rings or a tiered approach?

When an inspection triggers into the expansion, there is a unit-specific need to evaluate the results against the differences in materials used for the different locations and results from other unit inspections. For example, if failures are noted in the Alloy A-286 LCB bolts, but the UTS bolts are made from Alloy X-750 and no failures of this bolting material has been observed at any of the operating B&W units, a justification not to expand to this location should be possible. In addition, should failures be noted in a heat of Alloy A-286 used at one unit, expansion into the other units may need to be considered.

However, one or more of the bolts with indications need to be removed for laboratory testing to confirm the IGSCC failure mechanism and stress analyses need to be performed for each of the expansion bolt locations. Thus, expansion is not to be considered carte-blanche without additional evaluation.

WCAP- 17096-NP December 2009 Revision 2

A-21 Baffle-to-Former Bolts Effect,, Expansion Examination, :Examination Item Applicability (Mechanism) Link Method/Frequency Coverage Core Barrel All plants Cracking Baffle-to-baffle Baseline volumetric 100% of Assembly (IASCC, IE, bolts examination (UT) no accessible IC/ISRC later than two bolts.

Baffle-to-former Fatigue/Wear, Core refueling outages bolts Overload) barrel-to-forme from the begiing See MRP-231 r bolts of the license Figure 3-2.

renewal period with subsequent examination after 10 to 15 additional years.

There is a potential for failure in the form of cracking of baffle-to-former bolting to occur mainly from irradiation-assisted stress corrosion cracking (IASCC), but also as a result of irradiation embrittlement, irradiation creep/stress relaxation (leading to fatigue and wear), or overload (from a prying effect). Past failure history exists with baffle-to-former bolt materials (Type 316CW and Type 347) and core barrel-to-former bolt materials (Alloy X-750 and Type 316Ti CW) in non-B&W-design units. Currently, there are no known failures with any of the bolts (Type 304) in the operating B&W-design units. CR-3 has observed what appear to possibly be failed baffle-to-baffle bolts, but confirmation has never been made.

Component Item Function The core barrel assembly consists of the core barrel cylinder, former plates, and baffle plates connected by bolted joints that include: (1) core barrel-to-former bolts (CF bolts), (2) baffle-to-former bolts (BF bolts),

and (3) baffle-to-baffle bolts (BB bolts). The core barrel assembly supports the fuel assemblies, lower grid, flow distributor, and incore instrument guide tubes. The baffle plates, former plates, and their joints (including BF bolts) do not have a core support function and are categorized as internals structures. The primary function of the baffle plates, former plates, and their bolted connections is to provide a flow envelope surrounding the core. Also, since they are bolted to the core barrel cylinder, the baffle plates and former plates will produce a small increase on the stiffness and natural frequencies of the core barrel assembly.

The CF and BF bolts have the function of maintaining structural integrity of the baffle and former portion of the structural assembly and thus of maintaining flow geometry during normal operation. For faulted events, a small number of the CF and BF bolts are needed to restrain the baffle so that a coolable core geometry is maintained.

As with BB and CF bolts, loss of BF bolts will also influence changes in the core bypass flow due to opening of baffle-to-baffle corner gaps.

WCAP- 17096-NP December 2009 Revision 2

A-22 Observable Effects:

Cracking of the bolts is the main concern. A volumetric examination (UT) of the bolts is to be performed to identify such cracking.

Mockups and qualification efforts exist from the CR-3 examination in 2005, but there are several bolt designs in the B&W units and additional effort would still need to be completed.

Possible Examination Outcomes:

Cracking is anticipated to occur at the head-to-shank area where the peak tensile stress exists (i.e., a SCF exists) and OE has shown them to crack at this location in the past, although it may also occur in the shank thread region where high tensile stress is possible too.

No relevant conditions identified Relevant conditions (i.e., crack-like indications, either completely cracked or partially cracked; or non-interpretable UT indications, such as no back wall reflection or multiple reflections with no crack-like indication that is most likely caused by a large or duplex grain size) are identified Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the BF bolts involves the following steps and inputs:

A global finite element model (FEM) is developed to evaluate bolt failures for use in developing the frequency for the I&E guidelines, acceptable failure pattern or numbers, and for use in preparing possible JCOs for the BF bolts

• Representative rejected BF bolts are to be removed for laboratory testing to confirm the UT inspection results and IGSCC mechanism, or to identify any other failure mechanism(s)

NOTE: One alternative to removing representative UT rejected bolts for laboratory examination is to re-inspect all bolts at the next refueling outage if continued operation for one additional fuel cycle can be supported by a technical evaluation. Other possible options such as replacement of a predefmed bolt pattern may also be pursued.

The following inputs are required:

- Failed or missing BF bolt locations are required for input

- Thermal input including gamma-heating for design (short-term) and long-term operating conditions WCAP- 17096-NP December 2009 Revision 2

A-23 Irradiated material property input as a function of aging (EFPY) of the core barrel assembly Applicable test data to establish stress and/or strain and fatigue strength limits of the BF bolts in addition to the licensing basis requirements Faulted load licensing basis requirements based on existing evaluations and modified as needed Input of existing fuel baffle jetting evaluations and their applicability for the core baffle assembly in degraded conditions and modified as needed to establish if gap displacements are necessary acceptance criteria Input core barrel motion due to turbulence induced vibration from applicable startup testing and analytical evaluations A probability of failure of the inaccessible core barrel-to-former bolts and external baffle-to-baffle bolts will need to be determined Appropriate structural evaluations are performed to demonstrate the above acceptance criteria If necessary, the existing model will be modified to be suitable for dynamic loadings such as imposed core barrel motion due to turbulence induced vibration Analytical efforts could be performed on a generic basis, although there are two designs (ONS-1 and TMI-1 are one design and the other five units are the second design).

Existing Documentation:

A FEM has been developed by the MRP Reactor Internals project that can be used in performing the evaluations

- A few bolts could be identified as failed (non-interpretable bolt UT signal; inaccessible bolt for UT; locking device observed to be missing, non-functional, or removed; partially cracked bolt; or completely cracked bolt) and shown to be acceptable with no further action needed

- A number of bolts (TBD) identified as failed (non-interpretable UT signal, partially cracked, or completely cracked) would initiate replacement activities and lead into the expansion category Past B&WOG work has determined the minimum number of bolts required for safe shutdown, but not for operation; i.e., no minimum bolt patterns have been determined WCAP- 17096-NP December 2009 Revision 2

A-24 No acceptable evaluation or analysis has been completed to date for determining a re-inspection schedule What observations trigger examination into the Expansion category?

Cracking observed in 5% (40) of the bolts or greater than 25% of the bolts on a single former plate Should it trigger expansion to all remaining bolt rings or a tiered approach?

When an inspection triggers into the expansion, an evaluation of the internal BB bolts shall be performed to determine whether to examine or replace the internal BB bolts. The evaluation may also include an evaluation of the external BB bolts and CF bolts for the purpose of determining whether to continue operation or further expand into replacement activities.

WCAP- 17096-NP December 2009 Revision 2

A-25 Baffle Plates Effect Expansion Examination Examination' Item Applicabilit (Mechanisn),, Link Method/Frequency Coverage Core Barrel All plants Cracking (JE), Core barrel Visual (VT-3) 100% of the Assembly including the cylinder examination during accessible surface detection of (including the next 10-year ISI. within 1 inch Baffle plates readily vertical and around each flow detectable circumferential Subsequent and bolt hole.

cracking in the baffle plates seam welds) exmntosnth 10-year ISI interval. See MRP-231 Former plates Figure 3-2.

Baffle plates are subject to irradiation embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the materials degraded fracture toughness, such a condition could potentially lead to cracking. There is no known history of cracking of baffle plate material in PWR reactor vessel internals applications.

Component Item Function Degradation of the baffle plates could result in increased core bypass flow and a reduction in margin to DNB, but would probably have a negligible effect on unit operations and would not be observed except by direct examination. The core barrel supports the fuel assemblies, lower grid, flow distributor, and incore instrument guide tubes. However, the baffle plates do not support any load. The primary function of the baffle plates during normal power operation is to provide a flow envelope for the core and, thereby limit core bypass flow.

The baffle plates therefore do not have a direct core support safety function; however, they do have a safety function to control bypass around the core during a loss-of-coolant-accident (LOCA).

Observable Effects:

A visual (VT-3) examination of the baffle plates is to be performed. Subsequent visual examinations are to be performed during the ASME Code B-N-3 10-year ISI activities.

If flaws are identified, follow-on examination by VT- 1, ET, or UT may need to be performed to characterize the length or both length and depth of observations.

The baffle plates are being examined to detect large surface cracks. The locations expected to be subjected to the highest tensile stresses are near the baffle bolt holes (baffle-to-former and baffle-to-baffle) and flow holes/slots. Examinations should also include areas near the HAZ of the baffle bolt locking devices where residual tensile stresses may exist.

Possible Examination Outcomes:

0 No relevant conditions identified 0 One or more areas are identified with minor (short) crack-like indications WCAP-17096-NP December 2009 Revision 2

A-26

• One or more areas are identified with large (long) crack-like indications

• One or more areas are identified with missing material Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the baffle plates involves the following steps and inputs:

Confirmation of required loading and combination requirements Determine the expected crack opening displacement (COD) for development of the inspection standard Perform a linear-elastic fracture mechanics (LEFM) evaluation to determine the critical crack size using the MRP-211 fracture toughness values

- A flaw handbook could also be developed

- Or, justify the existing calculations in MRP-210

• Perform a bypass analysis to justify that sufficient DNB exists in the degraded condition

• A VT-1, ET, or UT examination may be needed to further characterize the flaws or to determine if flaws are emanating from the location where missing material may be identified

• An operability evaluation to operate at least one cycle based on possible inspection results for the plates should be performed a An evaluation of the consequences of leaving cracked plates securely in place during an inspection or replacement campaign should be performed The general methodology to be used for acceptance criteria for the baffle plates will be development of an NDE inspection standard that contains examples of acceptable and unacceptable visual indications and mockups for the VT-3 inspection of cracking. Input information needed includes:

Identification of the most likely locations of cracking in the plates Identification of what visual examination indications are considered rejectable and would require additional examination and evaluation Analytical efforts could be performed on a generic basis. The NDE inspection standard could also be developed generically.

WCAP- 17096-NP December 2009 Revision 2

A-27 Existing Documentation:

• CLB loadings (normal and faulted condition) are available, but a records search may need to be performed to identify them

• No CGRs currently exist

- CGRs for BWR HWC can be assumed for initial studies What observations trigger examination into the Expansion category?

• Gross cracking (if confirmed) on two or more locations in the baffle plates shall trigger an evaluation of the inspectability of the accessible areas of the former plates and core barrel (particularly the core barrel upper flange-to-core barrel weld and upper HAZ area) using VT-3

• inspection

• In addition, an evaluation of the operability of the former plates and core barrel in degraded conditions shall be performed

• If a VT-3 examination is possible, it is required by completion of the next refueling outage December 2009 WCAP- 17096-NP WCAP-17096-NP December 2009 Revision 2

A-28 Locking Devices of Baffle-to-Former and Internal Baffle-to-Baffle Bolts Examination Effect- Expansion Method xamina o Item Applicability (Mechanism) Link Frequency C6overage Core Barrel All plants Cracking Locking Visual (VT-3) 100% of accessible Assembly (IASCC, IE, devices for the examination baffle-to-former Overload), external during the next and internal Locking devices, including the baffle-to-baffle 10-year ISI. baffle-to-baffle bolt including locking detection of bolts and core locking devices.

welds, of missing, barrel-to- Subsequent lcigdvcs baffle-to-former non-functional, former bolts examinations See MRP-231 bolts and internal or removed on the 10-year Figure 3-2.

baffle-to-baffle locking devices ISI interval.

bolts or welds There is a potential for irradiation-assisted stress corrosion cracking (IASCC) of the locking devices or welds for the baffle (baffle-to-former and internal baffle-to-baffle) bolting. There is also the potential for failure by overload for these locking devices and welds due to slip between the bolts and plates (see MRP-23 1). Past failure history exists with cracked and missing internal baffle-to-baffle bolt locking devices and cracked locking device welds in B&W-design units.

Component Item Function The locking devices and welds are not normally loaded unless the bolt is broken or loose. Loading of the locking devices and welds could also occur due to the slip between the bolt and plate. The locking devices and welds have no core support safety function.

Observable Effects:

A visual (VT-3) examination of the bolt locking devices and welds is to be performed.

Cracking of the locking devices or welds is the main concern and they are to be examined to identify if any are distorted, loose, broken, or missing.

Possible Examination Outcomes:

• No relevant conditions identified

• One or a few bolt locking devices (up to 1% or 11) are identified with damage or are missing

• More than 1% (or, 11) bolt locking devices are identified with damage or are missing Methodology and Data Requirements:

The generalmethodology to be used for acceptance criteria for the locking devices and welds will be development of an NDE inspection standard that contains examples of acceptable and unacceptable locking device or weld visual indications. The acceptance of locking devices and welds is evaluated in two ways: a) observations with "failed" bolts and b) observations with all bolts intact. Observations of WCAP- 17096-NP December 2009 Revision 2

A-29 damaged locking devices or welds with all bolts intact represent a condition very different from that of locking device or weld damage at a bolt location that is failed. In addition, a damaged or missing welded locking washer versus a welded locking pin or bar potentially represents different initiating phenomena that need to be evaluated.

If bolt failure is not obvious, additional UT examinations of the bolt or a technical justification for removal of the locking device and bolt may be necessary.

The NDE inspection standard for the locking devices and welds could be performed on a generic basis for the B&W units, although there are two designs (ONS- 1 and TMI- 1 are one design and the other five units are the second design).

Existing Documentation:

Damaged and missing locking devices and welds from the CR-3 visual examinations can be used for development of an inspection standard, although additional efforts for other locking devices and welds will also be required.

What observations trigger examination into the Expansion category?

Confirmed rejectable indications in greater than or equal to 1% (or, 11) of the baffle-to-former and internal BB bolt locking devices shall trigger an evaluation of the locking devices for the external baffle-to-baffle bolts and core barrel-to-former bolts for the purpose of determining continued operation or replacement WCAP- 17096-NP December 2009 Revision 2

A-30 Guide Block Dowel Welds Effect Expansion Examination Examination Item Applicability (Mechanism) Link Method/Frequency Coverage Lower Grid All plants Cracking Alloy X-750 Initial visual (VT-3) 100% of Assembly (SCC), dowel locking examination no later accessible including the welds to the than two refueling locking welds of Alloy X-750 detection of upper and lower outages from the the 24 dowel-to-guide separated or grid fuel beginning of the dowel-to-guide block welds missing assembly license renewal block welds.

locking welds, support pads period. See MRP-231 or missing Subsequent Figure 3-4.

dowels examinations on the 10-year ISI interval.

There is a potential for primary water stress corrosion cracking (PWSCC) of the dowel-to-guide block welds. No known OE exists for cracking in these welds either in B&W-design units or other vendor designs, but OE with cracking of Alloy 82/182 weld materials in PWRs does exist.

Component Item Function These welds serve as loose part prevention devices and are not structural. Small cracks in the locking weld are acceptable since the locking function can be maintained as long as any part of the weld is present.

The dowel-to-guide block welds have no core support safety function.

Observable Effects:

A visual (VT-3) examination of the dowel-to-guide block welds is to be performed.

Loss of the locking function of the weld is the main concern and therefore the dowel-to-guide block welds are to be examined to identify if any are separated or missing or if a dowel is missing.

Possible Examination Outcomes:

0 No relevant conditions exist 0 A single weld is observed to be damaged or partially missing 0 Several welds are observed to be damaged or partially missing 0 A single weld or dowel is missing 0 Several welds or dowels are missing 0 A guide block is misaligned or is missing 0 Several guide blocks are misaligned or are missing WCAP- 17096-NP December 2009 Revision 2

A-31 Methodologv and Data Reouirements:

The general methodology to be used for acceptance criteria for the dowel-to-guide block welds will be development of an NDE inspection standard that contains examples of acceptable and unacceptable dowel-to-guide block weld visual indications. The function of the weld can be maintained as long as a portion of the weld is in place. Significant cracking of the weld and subsequent loss of the dowel does not compromise the function of the guide block unless the bolt also fails.

The following items will be examined to establish VT-3 acceptance criteria and the technical justification:

0 Identify normal and faulted operating loads for the guide block dowels 0 Evaluate how many (if any) guide blocks are needed for operation

• Evaluate the consequences of leaving partially cracked locking welds securely in place during an inspection

• Identify the areas to be examined containing what is rejectable and requiring further evaluation a Develop repair strategies for leaving in place if secured from being a loose part or removal and replacement activities A UT examination of the guide block bolt or a technical justification for removal of the guide block may be necessary.

Analytical efforts could be performed on a generic basis for the B&W units. The NDE inspection standard could also be developed generically.

Existing Documentation:

Minimal information is currently available.

What observations trigger examination into the Exoansion categorv?

• Confirmed rejectable indications of two or more separated, cracked, or failed locking welds shall trigger a VT-3 examination of the expansion category locking welds by the next scheduled refueling outage WCAP-17096-NP December 2009 Revision 2

A-32 Incore Monitoring Instrumentation Guide Tube Spiders and Welds Effect Expansion Examination Examination:

, Item*, Applicability (Mechanism) Link Method/Frequency: QCoverage-',,

Incore All plants Cracking CRGT spacer Initial visual (VT-3) 100% of Monitoring (TE/IE), castings examination no later accessible top Instrumentation including the than two refueling surfaces of (IMI) Guide detection of Lower grid outages from the 52 spider castings Tube Assembly. fractured or fuel assembly beginning of the and welds to the missing support pad license renewal adjacent lower spider anrs itm: pad, period, grid rib section.

spiders or separation pad-to-rib of spider section welds, Subsequent See MRP-231 IMI guide tube arms from the Alloy X-750 examinations on the Figures 3-3 and spider-to-lower lower grid rib dowel, cap 10-year interval. 3-6.

grid rib section section at the screw, and welds weld their locking welds (Note: The pads, dowels, and cap screws are included because of TE/IE of the welds)

The IMI guide tube spiders and welds are subject to both thermal aging and irradiation embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the materials degraded fracture toughness, such a condition could potentially lead to cracking. There is currently no known history of cracking of CASS or weld material subjected to such embrittlement in PWR reactor vessel internals applications.

Component Item Function Degradation of the IMI guide tube spiders could result in misalignment of the IMI nozzles and subsequent insertion of the in-core monitoring instrumentation. The IMI guide tube spiders do not have a core support safety function.

Observable Effects A visual (VT-3) examination of the IMI guide tube spiders is to be performed. Subsequent visual examinations are to be performed during the ASME Code B-N-3 10-year ISI activities.

The IMI guide tube spiders are being examined to detect spider arms that do not align with the lower fuel assembly support pad center bolt. The location that potentially contains the highest tensile stresses is near the heat-affected-zone (HAZ) of the spider-to-lower grid rib section weld, which is not readily accessible.

WCAP- 17096-NP December 2009 Revision 2

A-33 Possible Examination Outcomes:

• No relevant conditions identified

• One or more areas are identified with a spider arm that is not aligned with the lower fuel assembly support pad center bolt or obvious separation or the spider arm from the lower grid rib section welded connection

• One or more areas are identified with a missing spider arm Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the IMI guide tube spiders involves the following steps and inputs:

• Prepare an analysis to show that one or more missing spider arms or a completely missing spider will not result in loss of function of the IMI guide tube

• A VT-1, ET, or UT examination may be needed to determine if flaws are emanating from the location where missing material may be identified The general methodology to be used for acceptance criteria for these component items will be development of an NDE inspection standard that contains examples of acceptable and unacceptable visual indications and mockups for the VT-3 inspection of cracking. Input information needed includes:

• Identification of the most likely locations of surface irregularities, such as fractured or missing spider arms or separation of spider arms from the lower grid rib section at the weld

• Identification of what visual examination indications are considered rejectable and would require additional examination and evaluation Analytical efforts could be performed on a generic basis for each of the B&W units. The NDE inspection standard could also be developed generically.

Existing Documentation:

• CLB loadings (normal and faulted condition) are available, but a records search may need to be performed to identify them What observations trigger examination into the Expansion category?

Confirmed evidence of misalignment, separation, gross damage, or a missing spider arm for two or more locations shall trigger VT-3 examination of 100% of the accessible surfaces at the 4 screw locations (at every 900) of the CRGT spacer castings and the lower grid assembly support pad items and it is required by completion of the next refueling outage.

WCAP- 17096-NP December 2009 Revision 2

A-34 Expansion Component Items Acceptance criteria methodology and data requirements for each of the Expansion component items are summarized in this section. A separate sub-section is provided for each component item using the following format:

• Expansion component item information extracted directly from Table 4-4 of MRP-227

• This information is in tabular form and contains the item name, unit applicability, failure effect, failure mechanism(s), expansion link(s), examination method, examination frequency, and examination coverage.

• Component item function(s), including whether or not it has a core support safety function

• Observable effect(s)

• Methodology for development of acceptance criteria

• Data requirements for development of acceptance criteria

• Existing documents (e.g., PWROG or AREVA)

WCAP- 17096-NP December 2009 Revision 2

A-35 Upper and Lower Grid Fuel Assembly Support Pad Dowel Welds Effect. Examination Examination Item 'Applicability (Mechanism) , Primary Link Method ,'Coverage Upper or Lower Grid All plants Cracking Alloy X-750 Visual (VT-3) 100% of Assembly (SCC), dowel-to-guide examination, accessible (except upper including the block welds dowel locking Alloy X-750 grid assembly detection of welds.

dowel-to-upper grid fuel at DB) separated or assembly support pad missing See MRP-231 welds or Alloy X-750 locking weld, Figure 3-6.

dowel-to-lower grid fuel or missing assembly support pad dowels welds There is a potential for primary water stress corrosion cracking (PWSCC) of the dowel-to-grid fuel assembly support pad (both upper and lower grids) welds. No known OE of cracking exists for these welds either in B&W-design units or other vendor designs, but OE with cracking of Alloy 82/182 weld materials in PWRs does exist.

Component Item Function These welds serve as loose part prevention devices and are not structural. Small cracks in the locking weld are acceptable since the locking function can be maintained as long as any part of the weld is present. The fuel assembly support pads serve as guidance for loading of the fuel into the core. Once the fuel assemblies are loaded into the core, the support pads no longer have any function.

Therefore, the dowel-to-grid fuel assembly support pad welds have no core support safety function.

Observable Effects:

A visual (VT-3) examination of the dowel-to-grid fuel assembly support pad (upper and lower grids) welds is to be performed.

Loss of the locking function of the weld is the main concern and therefore the dowel-to-grid fuel assembly support pad welds are to be examined to identify if any are separated or missing, if a dowel is missing, or the support pad is misaligned (clearly out of perpendicularity).

Possible Examination Outcomes:

0 No relevant conditions exist 0 A single weld is observed to be damaged or partially missing 0 Several welds are observed to be damaged or partially missing 0 A single weld or dowel is missing 0 Several welds or dowels are missing 0 A support pad is misaligned (clearly out of perpendicularity) or missing 0 Several support pads are misaligned (clearly out of perpendicularity) or missing WCAP- 17096-NP December 2009 Revision 2

A-36 Methodology and Data Requirements:

The general methodology to be used for acceptance criteria for the dowel-to-grid fuel assembly support welds will be development of an NDE inspection standard that contains examples of acceptable and unacceptable dowel-to-grid fuel assembly support pad welds visual indications. The function of the weld can be maintained as long as a portion of the weld is in place. Significant cracking of the weld and subsequent loss of the dowel does not compromise the function of the fuel assembly support pad unless the bolt also fails.

The following items will be examined to establish VT-3 acceptance criteria and the technical justification:

• Identify normal and faulted operating loads for the fuel assembly support pad dowels

• Evaluate the consequences of leaving partially cracked locking welds securely in place during an inspection

• Identify the areas to be examined containing what is rejectable and requiring further evaluation

• Develop repair strategies for leaving in place if secured from being a loose part or removal and replacement activities A UT examination of the fuel assembly support pad bolt or a technical justification for removal of the fuel assembly support pad may be necessary.

Analytical efforts could be performed on a generic basis for the applicable locations at each of the B&W units. The NDE inspection standard could also be developed generically.

Existing Documentation:

Minimal information is currently available.

December 2009 WCAP- 17096-NP WCAY- 17096-NP December 2009 Revision 2

A-37 Control Rod Guide Tube Spacer Castings Effct- Exmiation Ex'aminato Item Applicability (MechaniSm) Primary Link I 'Method J ' Coverage Control Rod All plants Cracking (TE), CSS cast Visual 100% of accessible Guide Tube including the outlet nozzle (VT-3) surfaces at the Assembly detection fractured of CSS vent examination. 4(teey9° screw locations spacers or valve discs (at every 900)

CRGT spacer castngs paces or(Limited castings missing screws IMI guide tube accessibility)e spiders See MRP-231 Figure 3-5.

CRGT spacer castings are subject to thermal aging embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the materials degraded fracture toughness, such a condition could potentially lead to cracking. There is no known history of cracking of CASS material in PWR reactor vessel internals applications.

Component Item Function Degradation of the spacer castings could result in degradation in the unit shutdown capability. The spacer castings do not have a core support safety function; however, they do have a safety function relative to control rod alignment, insertion and reactivity issues, and a stuck rod scenario.

Observable Effects A visual (VT-3) examination of the CRGT spacer castings is to be performed.

The spacer castings have limited accessibility from the top or bottom of the CRGT through a center free-path (once the plenum assembly is removed from the vessel). Examination at the quarter points where the threaded connections are present is possible. These lanes are not blocked by the rod guide tubes. The examination would look for cracking of the spacer surface or evidence that the spacer is not approximately centered. The threaded fasteners are welded to the OD of the pipe column so it is possible that a degraded threaded location would not be detected.

Possible Examination Outcomes:

No relevant conditions identified One or more areas are identified with a large crack like indication that would be a precursor to loosing a piece of material One or more areas are identified with missing material Evidence that the spacer is not centered WCAP- 17096-NP December 2009 Revision 2

A-38 Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the CRGT spacer castings involves the following step and input:

• Perform a reactivity analysis to determine the number of CRDMs that are required for shut down of the reactor A VT- 1, ET, or UT examination may be needed to determine if flaws are emanating from the location where missing material may be identified.

The general methodology to be used for acceptance criteria for these component items will be development of an NDE inspection standard that contains examples of acceptable and unacceptable visual indications and mockups for the VT-3 inspection of fractured spacers or missing screws.

Analytical efforts for reactivity analyses are dependent upon fuel loading and must be performed on a unit-specific basis. The NDE inspection standard could also be developed generically.

Existing Documentation:

• CLB loadings (normal and faulted condition) are not currently available or even possibly required for analyses, but could be easily developed

• Core design WCAP- 17096-NP December 2009 Revision 2

A-39 Upper Thermal Shield Bolts and Locking Devices ffect "Examination Examination,,,

Item .... Applicability "(Mechanism) Primary Link Method Coverage Core Barrel All plants Cracking UCB bolts Volumetric 100% of Assembly (SCC) examination accessible bolts.

Upper thermal shield (UT). See MRP-231 bolts (UTS) and their Visual (VT-3) Figure 3-7.

examination locking devices of bolt locking devices on the 10-year ISI interval.

There is a potential for intergranular stress corrosion cracking (IGSCC) of Alloy A-286 and Alloy X-750 bolting. Past B&W failure history exists with the original Alloy A-286 bolt materials in B&W-design units and with applications of Alloy X-750 material within the nuclear industry (in general). Currently, there are no known failures with any of the replacement bolts (Alloy A-286 or Alloy X-750) in the operating B&W-design units or with the original Alloy X-750 (installed at TMI-1 only) in service in the operating B&W-design units.

Component Item Function The UTS bolts fasten a split restraint and shim block to the core barrel. The UTS bolts do not have a core support safety function.

Observable Effects:

A volumetric examination (UT) of the bolts and a visual (VT-3) examination of the bolt locking devices Mockups and qualification efforts exist from the PWROG work (PA-MSC-350) and additional Duke Energy efforts in 2007-2008.

Cracking of the bolts is the main concern and the locking devices are to be examined to identify if any are distorted, loose, broken, or missing.

The PWROG work (PA-MSC-350) also evaluated the potential information that could be determined from only a visual examination of the bolt and locking devices (see AREVA document 51-9081184-001).

Possible Examination Outcomes:

Cracking is anticipated to occur at the head-to-shank area where the peak tensile stress exists (i.e., a SCF exists) and OE has shown them to crack at this location in the past, although it may also occur in the shank thread region where high tensile stress is possible too.

No relevant conditions identified WCAP- 17096-NP December 2009 Revision 2

A-40

• Relevant conditions (i.e., crack-like indications, either completely cracked or partially cracked; or non-interpretable UT indications, such as no back wall reflection or multiple reflections with no crack-like indication that is most likely caused by a large or duplex grain size) are identified

- One or a few bolts (exact number is unit-specific) are identified with relevant indications

- More than a few bolts (exact number is unit-specific) are identified with relevant indications Locking Devices

• No relevant conditions identified

• One or two are identified with damage or are missing

• More than two bolt locking devices are identified with damage or are missing Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the bolts involves the following steps and inputs if relevant conditions have been identified in the UTS bolts:

• A finite element model (FEM) is to be developed for the local geometry with contact conditions, pretension elements, loads and boundary conditions

• A thermal analysis is to be performed

- Determines bolt temperatures and temperature gradients for normal operating conditions

• A structural analysis is to be performed in which failed bolts are inactive

- Stress concentration factors are calculated to determine the peak stresses at the bolt head-to-shank fillet region under normal operating conditions

- Analysis is performed for all loads and load combinations required for an ASME evaluation (stress limits for threaded structural fasteners in subsection NG and Appendix F)

- An evaluation of joint stability (or, openness) is also to be performed

• Representative rejected UTS bolts are to be removed for laboratory testing to confirm the UT inspection results and IGSCC mechanism, or to identify any other failure mechanism(s)

NOTE: One alternative to removing representative UT rejected bolts for laboratory examination is to re-inspect all bolts at the next refueling outage if continued operation for one additional fuel cycle can be supported by a technical evaluation. Other possible options such as replacement of a predefmed bolt pattern may also be pursued.

• Based on the results of UTS bolt UT inspection and laboratory test results, perform an evaluation to assess future UTS bolt failure potential. The changes to the peak stress at the bolt WCAP- 17096-NP December 2009 Revision 2

A-41 head-to-shank fillet region as a result of the identified failures should be included for evaluation of increased susceptibility to SCC.

Incorporate the effect of future UTS bolt failure into the operability evaluation and re-inspection requirement The general methodology to be used for acceptance criteria for the locking devices will be development of an NDE inspection standard that contains examples of acceptable and unacceptable locking device visual indications. The acceptance of locking devices is evaluated in two ways: a) observations with "failed" bolts and b) observations with all bolts intact. Observations of damaged locking devices with all bolts intact represent a condition very different from that of locking device damage at a bolt location that is failed. In addition, a damaged or missing welded locking clip versus a crimpled locking cup potentially represents different initiating phenomena that need to be evaluated.

Analytical efforts for the UTS bolt failures could be performed on a generic basis for all units except TMI-1, although use of unit-specific loadings could reduce the conservatism for some units. The NDE inspection standard could also be developed generically.

Existing Documentation:

An NRC-accepted crack growth rate for Alloy A-286 or Alloy X-750 material is not currently available.

However, the PWROG project (PA-MSC-350) has identified some CGR data that is currently available for a feasibility study of a life assessment approach, if desired.

WCAP- 17096-NP December 2009 Revision 2

A-42 Surveillance Specimen Holder Tube Studs/Nuts or Bolts and Locking Devices A Effect, -Examination Examination, Item Applicability (Mechanism), Primary Link Method Coverage Core Barrel CR-3, DB Cracking UCB bolts Volumetric 100% of Assembly (SCC) examination accessible bolts.

Surveilanceor bolts (UT) of studs bolts.

Surveillance specimen holder tube (SSHT) Visual (VT-3) studs/nuts (CR-3) or examination bolts (DB) and their of bolt locking locking devices devices or nuts on the 10 year ISI interval.

There is a potential for intergranular stress corrosion cracking (IGSCC) of Alloy A-286 and Alloy X-750 bolting. Past B&W failure history exists with the original Alloy A-286 bolt materials in B&W-design units and with applications of Alloy X-750 material within the nuclear industry (in general). Currently, there are no known failures with any of the replacement bolts (Alloy A-286 or Alloy X-750) in the operating B&W-design units or with the original Alloy X-750 (installed at TMI-1 only) in service in the operating B&W-design units.

Component Item Function The SSHT bolts fasten the surveillance specimen holder tubes to the thermal shield. Failure would result in loosening or dropping of the holder tube to the bottom of the vessel. The SSHT bolts do not have a core support safety function.

Observable Effects:

A volumetric examination (UT) of the bolts and a visual (VT-3) examination of the bolt locking devices Mockups and qualification efforts exist from the PWROG work (PA-MSC-350) and additional Duke Energy efforts in 2007-2008.

Cracking of the bolts is the main concern and the locking devices are to be examined to identify if any are distorted, loose, broken, or missing.

The PWROG work (PA-MSC-350) also evaluated the potential information that could be determined from only a visual examination of the bolt and locking devices (see AREVA document 51-9081184-001).

WCAP- 17096-NP December 2009 Revision 2

A-43 Possible Examination Outcomes:

Cracking is anticipated to occur at the head-to-shank area of the bolt where the peak tensile stress exists (i.e., a SCF exists) and OE has shown them to crack at this location in the past, although it may also occur in the shank or stud thread region where high tensile stress is possible too.

• No relevant conditions identified

• Relevant conditions (i.e., crack-like indications, either completely cracked or partially cracked; or non-interpretable UT indications, such as no back wall reflection or multiple reflections with no crack-like indication that is most likely caused by a large or duplex grain size) are identified

- One or a few studs/nuts or bolts (exact number is unit-specific) are identified with relevant indications

- More than a few studs/nuts or bolts (exact number is unit-specific) are identified with relevant indications Locking Devices

• No relevant conditions identified

• One or two are identified with damage or are missing

• More than two bolt locking devices are identified with damage or are missing Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the studs/nuts or bolts involves the following steps and inputs if relevant conditions have been identified in the SSHT studs/nuts or bolts:

A thermal analysis is to be performed

- Determines bolt temperatures and temperature gradients for normal operating conditions A structural analysis is to be performed in which failed studs/nuts or bolts are inactive

- Stress concentration factors are calculated to determine the peak stresses at the bolt head-to-shank fillet region or stud/nut thread region under normal operating conditions

- Analysis is performed for all loads and load combinations required for an ASME evaluation (stress limits for threaded structural fasteners in subsection NG and Appendix F)

- An evaluation of joint stability (or, openness) is also to be performed Representative rejected SSHT studs/nuts or bolts are to be removed for laboratory testing to confirm the UT inspection results and IGSCC mechanism, or to identify any other failure mechanism(s)

WCAP- 17096-NP December 2009 Revision 2

A-44 NOTE: One alternative to removing representative UT rejected bolts for laboratory examination is to re-inspect all bolts at the next refueling outage if continued operation for one additional fuel cycle can be supported by a technical evaluation. Other possible options such as replacement of a predefmed bolt pattern may also be pursued.

Based on the results of SSHT bolt UT inspection and laboratory test results, perform an evaluation to assess future SSHT stud/nut or bolt failure potential. The changes to the peak stress at the bolt head-to-shank fillet region or stud/nut thread region as a result of the identified failures should be included for evaluation of increased susceptibility to SCC.

Incorporate the effect of future SSHT stud/nut or bolt failure into the operability evaluation and re-inspection requirement The general methodology to be used for acceptance criteria for the locking devices will be development of an NDE inspection standard that contains examples of acceptable and unacceptable locking device visual indications. The acceptance of locking devices is evaluated in two ways: a) observations with "failed" studs/nuts or bolts and b) observations with all studs/nuts or bolts intact. Observations of damaged locking devices with all studs/nuts or bolts intact represent a condition very different from that of locking device damage at a stud/nut or bolt location that is failed. In addition, a damaged or missing welded locking clip versus a crimpled locking cup potentially represents different initiating phenomena that need to be evaluated.

Since there are only two units, and one has studs/nuts and the other has bolts, unit-specific analyses are required.

Existing Documentation:

An NRC-accepted crack growth rate for Alloy A-286 or Alloy X-750 material is not currently available.

However, the PWROG project (PA-MSC-350) has identified some CGR data that is currently available for a feasibility study of a life assessment approach, if desired.

WCAP- 17096-NP December 2009 Revision 2

A-45 Core Barrel Cylinder

'Examination Examination'-'

Item Applicability (Mechanism) Link Method Coverage Core Barrel All plants Cracking Baffle plates Justify by Inaccessible.

Assembly (IE), evaluation or by See MRP-231 Core barrel readilyFiue32 including replacement. Figure 3-2.

detectable cylinder (including ctable vertical and cracking circumferential seam welds)

The core barrel cylinders and welds are subject to irradiation embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the materials degraded fracture toughness, such a condition could potentially lead to cracking. There is no known history of OE for cracking of core barrel cylinder and weld material in PWR reactor vessel internals applications.

Comnonent Item Function Degradation of the core barrel cylinders and welds could result in increased core bypass flow and a reduction in margin to DNB, but would probably have a negligible effect on unit operations and would not be observed except by direct examination. The core barrel supports the fuel assemblies, lower grid, flow distributor, and incore instrument guide tubes. The primary function of the core barrel cylinders and welds during normal power operation is to provide a flow envelope for the core and, thereby limit core bypass flow.

The core barrel cylinders and welds therefore do not have a direct core support safety function; however, they do have a safety function to control bypass around the core during a loss-of-coolant-accident (LOCA).

Observable Effects:

The core barrel cylinders and welds are mostly inaccessible without disassembly. Therefore, no examination is currently required in MRP-227.

The core barrel upper flange-to-core barrel wed and upper HAZ area is partially accessible and could potentially be VT-3 examined.

Nothing is being examined at this time.

WCAP- 17096-NP December 2009 Revision 2

A-46 Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the core barrel cylinders and welds involves the following steps and inputs:

• Confirmation of required loading and combination requirements

• Perform a linear-elastic fracture mechanics (LEFM) evaluation to determine the critical crack size using the MRP-211 fracture toughness values

- A flaw handbook could also be developed

- Or, justify the existing calculations in MRP-210

• Perform a bypass analysis to justify that sufficient DNB exists in the degraded condition

• An operability evaluation to operate at least one cycle based on possible degradation of the core barrel cylinders and welds should be performed

• An evaluation of the consequences of leaving cracked core barrel cylinders and welds in place during an inspection or replacement campaign should be performed Analytical efforts could be performed on a generic basis.

Existing Documentation:

• CLB loadings (normal and faulted condition) are available, but a records search may need to be performed to identify them

• No CGRs currently exist

- CGRs for BWR HWC can be assumed for initial studies WCAP- 17096-NP December 2009 Revision 2

A-47 Former Plates Effect Primary Examination Examination Item Applicability (MUchanism) Lik Method . Coverage Core Barrel All plants Cracking (IE), Baffle plates Justify by Inaccessible.

Assembly including readily evaluation or by See MRP-231 detectable replacement. Figure 3-2.

Former plates cracking Former plates are subject to irradiation embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the materials degraded fracture toughness; such a condition could potentially lead to cracking. There is no known history of OE for cracking of former plate material in PWR reactor vessel internals applications.

Component Item Function The former plates do not have a direct core support safety function; however, they do have a safety function to help maintain the structural integrity of the core barrel assembly during operating conditions.

Observable Effects:

The former plates are mostly inaccessible without disassembly. Therefore, no examination is currently required in MRP-227.

The former plates are partially accessible through openings in the core barrel assembly and could potentially be VT-3 examined.

Nothing is required for examination at this time.

Methodology and Data Reguirements:

The general analytical methodology to be used for acceptance criteria for the former plates involves the following steps and inputs:

• Confirmation of required loading and combination requirements

• Perform a linear-elastic fracture mechanics (LEFM) evaluation to determine the critical crack size using the MRP-211 fracture toughness values

- A flaw handbook could also be developed

- Or, justify the existing calculations in MRP-2 10 WCAP- 17096-NP December 2009 Revision 2

A-48

• An operability evaluation to operate at least one cycle based on possible degradation of the former plates should be performed

• An evaluation of the consequences of leaving cracked former plates in place should be performed Analytical efforts could be performed on a generic basis.

Existing Documentation:

• CLB loadings (normal and faulted condition) are available, but a records search may need to be performed to identify them

• No CGRs currently exist

- CGRs for BWR HWC can be assumed for initial studies WCAP- 17096-NP December 2009 Revision 2

A-49 Baffle-to-Baffle Bolts

, Effect." Examination Examination

.tem A lcability (Mechanism) Primary Link Cverage CMethod Core Barrel All plants Cracking Baffle-to-former Internal Acceptable Assembly (IASCC, IE, bolts baffle-to-baffle examination IC/ISR/ bolts: technique not Baffle-to-baffle bolts Fatigue/Wear, currently available.

bolts Overload) No examination requirements, See MRP-231 justify by Figure 3-2.

evaluation or by replacement.

External Inaccessible.

baffle-to-baffle bolts:See MRP-231 Figure 3-2.

No examination requirements, justify by evaluation or by replacement.

There is a potential for failure in the form of cracking of baffle-to-baffle bolting to occur mainly from irradiation-assisted stress corrosion cracking (IASCC), but also as a result of irradiation embrittlement, irradiation creep/stress relaxation (leading to fatigue and wear), or overload (from a prying effect). Past failure history exists with baffle-to-former bolt materials (Type 316CW and Type 347) and core barrel-to-former bolt materials (Alloy X-750 and Type 316Ti CW) in non-B&W-design units. Currently, there are no known failures with any of the bolts (Type 304) in the operating B&W-design units. CR-3 has observed what appear to possibly be failed internal baffle-to-baffle bolts, but confirmation has never been made.

Comnonent Item Function The core barrel assembly consists of the core barrel cylinder, former plates, and baffle plates connected by bolted joints that include: (1) core barrel-to-former bolts (CF bolts), (2) baffle-to-former bolts (BF bolts),

and (3) baffle-to-baffle bolts (BB bolts). The core barrel assembly supports the fuel assemblies, lower grid, flow distributor, and incore instrument guide tubes. The baffle plates, former plates, and their joints (including BF bolts) do not have a core support function and are categorized as internals structures. The primary function of the baffle plates, former plates, and their bolted connections is to provide a flow envelope surrounding the core. Also, since they are bolted to the core barrel cylinder, the baffle plates and former plates will produce a small increase on the stiffness and natural frequencies of the core barrel assembly.

The CF and BF bolts have the function of maintaining structural integrity of the baffle and former portion of the structural assembly and thus of maintaining flow geometry during normal operation. For faulted events, a small number of the CF and BF bolts are needed to restrain the baffle so that a coolable core WCAP- 17096-NP December 2009 Revision 2

A-50 geometry is maintained. The BB bolts are not required for these functions but rather serve to minimize gaps between baffle plates. The BB bolts therefore do not have a core support safety function.

BB bolts are divided into two groups; those BB bolts on internal corners receive neutron fluence that is much higher than those on external corners. The two groups also differ in accessibility for inspection.

Observable Effects:

Cracking of the bolts is the main concern.

No examinations are required at this time in MRP-227.

Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the BB bolts involves the following steps and inputs:

• A global FEM model is developed to evaluate failures for use in developing the frequency for the I&E guidelines, acceptable failure pattern or numbers, and for use in preparing possible JCOs for the BF and CF bolts

- Evaluations for these bolt locations will consider BB bolts to be failed and structurally inactive

- No specific, pattern will need to be evaluated as the BB bolts do not perform any core support function, nor are they required to maintain the geometry of the core cavity

- In addition, no specific acceptance criteria are required for BB bolts

• A hydraulic analysis for evaluation of changes in jet momentum flux due to changes in gaps will be performed to assess changes in jetting and possible fuel failures Analytical efforts could be performed on all bolt locations on a generic basis.

Existing Documentation:

• A FEM has been developed by the MRP Reactor Internals project to evaluate failures that can be used in evaluating an acceptable failure pattern or number of failed bolts allowed for continued operation, and for use in preparing a JCO WCAP- 17096-NP December 2009 Revision 2

A-51 Core Barrel-to-Former Bolts Effect Examination Examination Item Applicability , (Mechanism)ý Prim~airy Link Method Coverage Core Barrel All plants Cracking (IASCC, Baffle-to-former No examination Inaccessible.

Assembly IE, IC/ISR/ bolts requirements, See MRP-231 Core Fatigue/Wear, justify by Fee Figure 3-2.

3-2.

barrel-to-former Overload) evaluation or by bolts IrItIrIrreplacement.

There is a potential for failure in the form of cracking of core barrel-to-former bolting to occur mainly from irradiation-assisted stress corrosion cracking (IASCC), but also as a result of irradiation embrittlement, irradiation creep/stress relaxation (leading to fatigue and wear), or overload (from a prying effect). Past failure history exists with baffle-to-former bolt materials (Type 316CW and Type 347) and core barrel-to-former bolt materials (Alloy X-750 and Type 316Ti CW) in non-B&W-design units.

Currently, there are no known failures with any of the bolts (Type 304) in the operating B&W-design units. CR-3 has observed what appear to possibly be failed baffle-to-baffle bolts, but confirmation has never been made.

Component Item Function The core barrel assembly consists of the core barrel cylinder, former plates, and baffle plates connected by bolted joints that include: (1) core barrel-to-former bolts (CF bolts), (2) baffle-to-former bolts (BF bolts),

and (3) baffle-to-baffle bolts (BB bolts). The core barrel assembly supports the fuel assemblies, lower grid, flow distributor, and incore instrument guide tubes. The baffle plates, former plates, and their joints (including BF bolts) do not have a core support function and are categorized as internals structures. The primary function of the baffle plates, former plates, and their bolted connections is to provide a flow envelope surrounding the core. Also, since they are bolted to the core barrel cylinder, the baffle plates and former plates will produce a small increase on the stiffness and natural frequencies of the core barrel assembly.

The CF and BF bolts have the function of maintaining structural integrity of the baffle and former portion of the structural assembly and thus of maintaining flow geometry during normal operation. For faulted events, a small number of the CF and BF bolts are needed to restrain the baffle so that a coolable core geometry is maintained.

Observable Effects:

Cracking of the bolts is the main concern.

No examinations are required at this time in MRP-227.

WCAP- 17096-NP December 2009 Revision 2

A-52 Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the CF bolts involves the following steps and inputs:

A global FEM model is developed to evaluate failures for use in developing the frequency for the I&E guidelines, acceptable failure pattern or numbers, and for use in preparing possible JCOs for the BF or CF bolts and this model can be used to evaluate the need for each of these two bolt locations The following inputs are required:

- Failed or missing CF bolt locations are required for input

- Thermal input including gamma-heating for design (short-term) and long-term operating conditions

- Irradiated material property input as a function of aging (EFPY) of the core barrel assembly

- Applicable test data to establish stress and/or strain and fatigue strength limits of the CF bolts in addition to the licensing basis requirements

- Faulted load licensing basis requirements based on existing evaluations and modified as needed

- Acceptable baffle displacements, changes in flow slot or gaps, and/or changes in natural frequency of the baffle/former structure, as applicable to the BF and BB bolts Appropriate structural evaluations are performed to demonstrate the above acceptance criteria If necessary, the existing model will be modified to be suitable for dynamic loadings such as imposed core barrel motion due to turbulence induced vibration Analytical efforts could be performed on all bolt locations on a generic basis.

Existing Documentation:

• A FEM has been developed by the MRP Reactor Internals project to evaluate failures that can be used in evaluating an acceptable failure pattern or number of failed bolts allowed for continued operation, and for use in preparing a JCO

- The FEM model for the MRP project has not been applied to dynamic loadings Past efforts however have assumed all CF bolts to be intact No acceptable evaluation or analysis has been completed to date for determining a re-inspection schedule WCAP- 17096-NP December 2009 Revision 2

A-53 Locking Devices for External Baffle-to-Baffle and Core Barrel-to-Former Bolts Effect,,, -  ;,Examination' Examination, Item- Applicabili chanism) Primary Link Method .. Cerag Core Barrel All plants Cracking Locking devices, Justify by Inaccessible.

Assembly (IASCC, IE) including locking evaluation or by welds, of replacement. See MRP-231 Locking devices, baffle-to-former Figure 3-2.

including locking bolts or internal welds, for the baffle-to-baffle external bolts baffle-to-baffle bolts and core barrel-to-former bolts There is a potential for irradiation-assisted stress corrosion cracking (IASCC) of the locking devices or welds for the external baffle-to-baffle and core barrel-to-former bolting. There is also the potential for failure by overload for these locking devices and welds due to slip between the bolts and plates (see MRP-23 1). Past failure history exists with cracked and missing internal baffle-to-baffle bolt locking devices and cracked locking device welds in B&W-design units.

Component Item Function The locking devices and welds are not normally loaded unless the bolt is broken or loose. Loading of the locking devices and welds could also occur due to the slip between the bolt and plate. The locking devices and welds have no core support safety function.

Observable Effects:

Items are inaccessible and no known technique is available other than disassembly of the core barrel assembly for a visual examination; therefore, no examinations are required at this time in MRP-227.

Cracking of the locking devices or welds is the concern.

Methodology and Data Requirements:

Locking device failure in itself is not a safety concern and an assessment can be prepared stating this as such. Failure of the bolting locations is of more concern and is covered in the bolting summary pages.

Analytical efforts could be performed on a generic basis for the B&W units.

Existing Documentation:

Nothing is available at this time.

WCAP- 17096-NP December 2009 Revision 2

A-54 Lower Grid Fuel Assembly Support Pad Items Effect Examination Examination Itemrn Applicabilit. (Mechanism) Primary Link Method Coverage Lower Grid All plants Cracking (IE), IMI guide tube Visual (VT-3) 100% of accessible Assembly including of detection the spiders examination, pads, dowels, and cap screws, and Lower grid fuel separated or Spider-to-lower associated welds.

assembly support missing welds, grid rib section pad items: pad, missing welds See MRP-231 pad-to-rib section support pads, Figure 3-6.

welds, dowels, cap Alloy X-750 screws and dowel, cap screw, locking welds, and their locking or welds misalignment (Note: The pads, of the support dowels, and cap pads.

screws are included because of TE/IE of the welds)

Lower grid fuel assembly support pad items are mostly subject to irradiation embrittlement, with some also susceptible to thermal aging and irradiation embrittlement, which if a flaw would be present and they are subjected to loading that exceeds the materials degraded fracture toughness, such a condition could potentially lead to cracking. There is no known history of OE for cracking of these materials in PWR reactor vessel internals applications.

Component Item Function These welds serve as loose part prevention devices and are not structural. Small cracks in the locking weld are acceptable since the locking function can be maintained as long as any part of the weld is present. The fuel assembly support pads serve as guidance for loading of the fuel into the core. Once the fuel assemblies are loaded into the core, the support pads no longer have any function.

Therefore, the lower grid fuel assembly support pad items have no core support safety function.

Observable Effects A visual (VT-3) examination of the lower grid fuel assembly support pad items is to be performed.

Loss of the lower grid fuel assembly support pad is the main concern and therefore the various items are to be examined to identify if any are separated or missing, if a dowel is missing, or the support pad is misaligned (clearly out of perpendicularity).

WCAP- 17096-NP December 2009 Revision 2

A-55 Possible Examination Outcomes:

• No relevant conditions exist a A single weld is separated or missing

• Several welds are separated or missing a A single dowel is missing a Several dowels are missing

• A support pad is misaligned (clearly out of perpendicularity) or missing

• Several support pads are misaligned (clearly out of perpendicularity) or missing Methodology and Data Reguirements:

The general methodology to be used for acceptance criteria for the lower grid fuel assembly support items will be development of an NDE inspection standard that contains examples of acceptable and unacceptable lower grid fuel assembly support pad item visual indications. The function of the support pad can be maintained as long as a portion of any of the welds is in place. Significant cracking of the welds and subsequent loss of the dowel does not compromise the function of the fuel assembly support pad unless the screw also fails.

The following items will be examined to establish VT-3 acceptance criteria and the technical justification:

Identify normal and faulted operating loads for the fuel assembly support pad dowels Evaluate the consequences of leaving partially cracked locking welds securely in place during an inspection Identify the areas to be examined containing what is rejectable and requiring further evaluation Develop repair strategies for leaving in place if secured from being a loose part or removal and replacement activities A UT examination of the fuel assembly support pad screw or a technical justification for removal of the fuel assembly support pad may be necessary.

Analytical efforts could be performed on a generic basis for the B&W units. The NDE inspection standard could also be developed generically.

Existing Documentation:

Minimal information is currently available.

WCAP- 17096-NP December 2009 Revision 2

A-56 Lower Grid Shock Pad Bolts and Locking Devices Effect Examination Examination Item Applicability '(Mechanism) Primary Link Method. Coverage Lower Grid TMI-1 Cracking UCB bolts Volumetric 100% of Assembly (SCC) examination accessible bolts.

Lower grid shock pad (UT). See MRP-231 bolts and their Visual (VT-3) Figure 3-4 locking devices examination of bolt locking devices on the 10-year ISI interval.

There is a potential for intergranular stress corrosion cracking (IGSCC) of Alloy A-286 and Alloy X-750 bolting. Past B&W failure history exists with the original Alloy A-286 bolt materials in B&W-design units and with applications of Alloy X-750 material within the nuclear industry (in general). Currently, there are no known failures with any of the replacement bolts (Alloy A-286 or Alloy X-750) in the operating B&W-design units or with the original Alloy X-750 (installed at TMI-1 only) in service in the operating B&W-design units.

Component Item Function The function of the lower grid shock pad bolts is to fasten the shock pads to the lower grid assembly.

Shock pads must be in place to carry accidental core drop loads. The bolts do not function to carry the core drop load, but serve to hold the pad in place. Each shock pad is held by two bolts. At least one must be intact on each shock pad to prevent a loose part.

The shock pad bolts are also part of the joint between the lower end of the thermal shield cylinder and the lower grid assembly. At TMI- 1, these bolts are fabricated from Alloy X-750 material and are designed to hold the shock pad in place and engage the lower thermal shield too. Hence, the shock pad bolts at TMI- 1 also function as LTS bolts. This is a unique design feature not shared by the other B&W units. The lower thermal shield joint acts as a restraint to vertical and rotational motion of the bottom of the thermal shield, but does not act as a direct core support component and therefore does not have a core support safety function. Evaluation of the lower thermal shield joint is based on 96 LTS bolts with no credit taken for strength of the shock pad bolts. Therefore, the only function to be maintained by the shock pad bolts is to keep the shock pads in place.

Observable Effects:

A volumetric examination (UT) of the bolts and a visual (VT-3) examination of the bolt locking devices WCAP- 17096-NP December 2009 Revision 2

A-57 Mockups and qualification efforts exist from the PWROG work (PA-MSC-350) and additional Duke Energy efforts in 2007-2008, Cracking of the bolts is the main concern and the locking devices are to be examined to identify if any are distorted, loose, broken, or missing.

The PWROG work (PA-MSC-350) also evaluated the potential information that could be determined from only a visual examination of the bolt and locking devices (see AREVA document 51-9081184-001).

Cracking is anticipated to occur at the head-to-shank area where the peak tensile stress exists (i.e., a SCF exists) and OE has shown them to crack at this location in the past, although it may also occur in the shank thread region where high tensile stress is possible too.

No relevant conditions identified Relevant conditions (i.e., crack-like indications, either completely cracked or partially cracked; or non-interpretable UT indications, such as no back wall reflection or multiple reflections with no crack-like indication that is most likely caused by a large or duplex grain size) are identified

- If one bolt is missing on a shock pad, it is regarded as relevant since there is no redundancy to hold the pad in place.

- If two bolts on any individual shock pad are identified with relevant conditions, the shock pad could become a loose part and may not be in place in the event of a core drop accident Locking Devices

• No relevant conditions identified

• One or two are identified with damage or are missing

• More than two bolt locking devices are identified with damage or are missing Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the bolts involves the following steps and inputs if relevant conditions have been identified in the lower grid shock pad bolts:

If two bolts on any individual shock pad are identified with relevant indications, it is an indication that the shock pad may become loose and will not be in place to carry a core drop. A structural evaluation is to be performed to determine if remaining pads can carry the core drop load or if the load can be carried without the shock pad in place If one of two bolts has an indication, an analysis is to be performed to assess loads on the remaining bolt and its potential for future failure

- Loads on this bolt will include those evaluated as part of modeling of the lower thermal shield joint as described for LTS bolts WCAP- 17096-NP December 2009 Revision 2

A-58 Stress concentration factors are calculated to determine the peak stresses at the bolt head-to-shank fillet region under normal operating conditions Structural evaluation will be performed to determine peak stress in remaining shock pad bolts for use in assessing potential for future bolt failure Representative rejected lower grid shock pad bolts are to be removed for laboratory testing to confirm the UT inspection results and IGSCC mechanism, or to identify any other failure mechanism(s)

NOTE: One alternative to removing representative UT rejected bolts for laboratory examination is to re-inspect all bolts at the next refueling outage if continued operation for one additional fuel cycle can be supported by a technical evaluation. Other possible options such as replacement of a predefined bolt pattern may also be pursued.

Based on the results of lower grid shock pad bolt UT inspection and laboratory test results, perform an evaluation to assess future lower grid shock pad bolt failure potential. The changes to the peak stress at the bolt head-to-shank fillet region as a result of the identified failures should be included for evaluation of increased susceptibility to SCC.

Incorporate the effect of future lower grid shock pad bolt failure into the operability evaluation and re-inspection requirement The general methodology to be used for acceptance criteria for the locking devices will be development of an NDE inspection standard that contains examples of acceptable and unacceptable locking device visual indications. The acceptance of locking devices is evaluated in two ways: a) observations with "failed" bolts and b) observations with all bolts intact. Observations of damaged locking devices with all bolts intact represent a condition very different from that of locking device damage at a bolt location that is failed. In addition, a damaged or missing welded locking clip versus a crimpled locking cup potentially represents different initiating phenomena that need to be evaluated.

A TMI unit-specific analytical effort is required.

Existing Documentation:

An NRC-accepted crack growth rate for Alloy A-286 or Alloy X-750 material is not currently available.

However, the PWROG project (PA-MSC-350) has identified some CGR data that is currently available for a feasibility study of a life assessment approach, if desired.

WCAP- 17096-NP December 2009 Revision 2

A-59 Lower Thermal Shield Bolts and Locking Devices

• . .* ,- : Effect Pr m r .' ....... . .':'*: :  :

E..et . Primary Examination Examination Item Applicability (Mechanism) Link Method Coverage.

Lower Grid All plants Cracking UCB bolts Volumetric 100% of Assembly (SCC) examination (UT) accessible bolts.

Lower thermal of studs or bolts. See MRP-231 shield studs/nuts or Visual (VT-3) Figure 3-8.

bolts (LTS) and examination of their bolt locking devices or nuts locking devices on the 10-year ISI interval.

There is a potential for intergranular stress corrosion cracking (IGSCC) of Alloy A-286 and Alloy X-750 bolting. Past B&W failure history exists with the original Alloy A-286 bolt materials in B&W-design units and with applications of Alloy X-750 material within the nuclear industry (in general). Currently, there are no known failures with any of the replacement bolts (Alloy A-286 or Alloy X-750) in the operating B&W-design units or with the original Alloy X-750 (installed at TMI-1 only) in service in the operating B&W-design units.

Component Item Function The LTS bolts fasten the thermal shield cylinder to the lower grid assembly. The LTS joint acts as a restraint to vertical and rotational motion of the bottom of the thermal shield, but does not act as a direct core support component and therefore does not have a core support safety function.

Observable Effects:

A volumetric examination (UT) of the studs/nuts or bolts and a visual (VT-3) examination of the stud/nut or bolt locking devices Mockups and qualification efforts exist from the PWROG work (PA-MSC-350) and additional Duke Energy efforts in 2007-2008.

Cracking of the bolts is the main concern and the locking devices are to be examined to identify if any are distorted, loose, broken, or missing.

The PWROG work (PA-MSC-350) also evaluated the potential information that could be determined from only a visual examination of the bolt and locking devices (see AREVA document 51-9081184-001).

WCAP- 17096-NP December 2009 Revision 2

A-60 Possible Examination Outcomes:

Cracking is anticipated to occur at the head-to-shank area of the bolt where the peak tensile stress exists (i.e., a SCF exists) and OE has shown them to crack at this location in the past, although it may also occur in the shank or stud thread region where high tensile stress is possible too.

• No relevant conditions identified

• Relevant conditions (i.e., crack-like indications, either completely cracked or partially cracked; or non-interpretable UT indications, such as no back wall reflection or multiple reflections with no crack-like indication that is most likely caused by a large or duplex grain size) are identified

- One or a few studs/nuts or bolts (exact number is unit-specific) are identified with relevant indications

- More than a few studs/nuts or bolts (exact number is unit-specific) are identified with relevant indications Locking Devices

• No relevant conditions identified

• One or two are identified with damage or are missing

• More than two bolt locking devices are identified with damage or are missing Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the studs/nuts or bolts involves the following steps and inputs if relevant conditions have been identified in the LTS studs/nuts or bolts:

• A finite element model (FEM) is to be developed for the local geometry with contact conditions, pretension elements, loads and boundary conditions

• A thermal analysis is to be performed

- Determines bolt temperatures and temperature gradients for normal operating conditions

• A structural analysis is to be performed in which failed studs/nuts or bolts are inactive

- Stress concentration factors are calculated to determine the peak stresses at the bolt head-to-shank fillet region or stud/nut thread region under normal operating conditions

- Analysis is performed for all loads and load combinations required for an ASME evaluation (stress limits for threaded structural fasteners in subsection NG and Appendix F)

- An evaluation of joint stability (or, openness) is also to be performed WCAP- 17096-NP December 2009 Revision 2

A-61 Representative rejected LTS studs/nuts or bolts are to be removed for laboratory testing to confirm the UT inspection results and IGSCC mechanism, or to identify any other failure mechanism(s)

NOTE: One alternative to removing representative UT rejected bolts for laboratory examination is to re-inspect all bolts at the next refueling outage if continued operation for one additional fuel cycle can be supported by a technical evaluation. Other possible options such as replacement of a predefmed bolt pattern may also be pursued.

Based on the results of LTS bolt UT inspection and laboratory test results, perform an evaluation to assess future LTS stud/nut or bolt failure potential. The changes to the peak stress at the bolt head-to-shank fillet region or stud/nut thread region as a result of the identified failures should be included for evaluation of increased susceptibility to SCC.

Incorporate the effect of future LTS stud/nut or bolt failure into the operability evaluation and re-inspection requirement The general methodology to be used for acceptance criteria for the locking devices will be development of an NDE inspection standard that contains examples of acceptable and unacceptable locking device visual indications. The acceptance of locking devices is evaluated in two ways: a) observations with "failed" studs/nuts or bolts and b) observations with all studs/nuts or bolts intact. Observations of damaged locking devices with all studs/nuts or bolts intact represent a condition very different from that of locking device damage at a stud/nut or bolt location that is failed. In addition, a damaged or missing welded locking clip versus a crimpled locking cup potentially represents different initiating phenomena that need to be evaluated.

Due to the variations in stud/nut or bolt materials used and loadings among the units, unit-specific analyses are required. The NDE inspection standard could be developed on a generic basis.

Existing Documentation:

An NRC-accepted crack growth rate for Alloy A-286 or Alloy X-750 material is not currently available.

However, the PWROG project (PA-MSC-350) has identified some CGR data that is currently available for a feasibility study of a life assessment approach, if desired.

WCAP- 17096-NP December 2009 Revision 2

A-62 Flow Distributor Bolts and Locking Devices Effect Examination Examinhation Item Applicability (Mechanism) Primary Link Method Coverage Flow Distributor All plants Cracking UCB bolts Volumetric 100% of Assembly (SCC) examination accessible bolts.

Flow distributor (UT). See MRP-231 bolts (FD) and Visual (VT-3) Figure 3-8.

their locking examination devices of bolt locking devices on the 10-year ISI interval.

There is a potential for intergranular stress corrosion cracking (IGSCC) of Alloy A-286 and Alloy X-750 bolting. Past B&W failure history exists with the original Alloy A-286 bolt materials in B&W-design units and with applications of Alloy X-750 material within the nuclear industry (in general). Currently, there are no known failures with any of the replacement bolts (Alloy A-286 or Alloy X-750) in the operating B&W-design units or with the original Alloy X-750 (installed at TMI- 1 only) in service in the operating B&W-design units.

Component Item Function The FD bolts fasten the flange on the flow distributor to the lower grid assembly. The joint also clamps the lower grid support plate in place between the bottom of the lower grid assembly and a ledge on the ID of the flow distributor. A clamp ring spans the gap between the bottom mating face of the lower grid assembly and the top of the support plate, providing the compressive force holding the support plate in place. The FD bolts do not have a core support safety function.

Observable Effects:

A volumetric examination (UT) of the bolts and a visual (VT-3) examination of the bolt locking devices Mockups and qualification efforts exist from the PWROG work (PA-MSC-350) and additional Duke Energy efforts in 2007-2008.

Cracking of the bolts is the main concern and the locking devices are to be examined to identify if any are distorted, loose, broken, or missing.

The PWROG work (PA-MSC-350) also evaluated the potential information that could be determined from only a visual examination of the bolt and locking devices (see AREVA document 51-9081184-001).

WCAP-17096-NP December 2009 Revision 2

A-63 Possible Examination Outcomes:

Bolts Cracking is anticipated to occur at the head-to-shank area where the peak tensile stress exists (i.e., a SCF exists) and OE has shown them to crack at this location in the past, although it may also occur in the shank thread region where high tensile stress is possible too.

• No relevant conditions identified

• Relevant conditions (i.e., crack-like indications, either completely cracked or partially cracked; or non-interpretable UT indications, such as no back wall reflection or multiple reflections with no crack-like indication that is most likely caused by a large or duplex grain size) are identified

- One or a few bolts (exact number is unit-specific) are identified with relevant indications

- More than a few bolts (exact number is unit-specific) are identified with relevant indications Locking Devices

• No relevant conditions identified

• One or two are identified with damage or are missing

• More than two bolt locking devices are identified with damage or are missing Methodology and Data Requirements:

The general analytical methodology to be used for acceptance criteria for the bolts involves the following steps and inputs if relevant conditions have been identified in the FD bolts:

• A finite element model (FEM) is to be developed for the local geometry with contact conditions, pretension elements, loads and boundary conditions

• A thermal analysis is to be performed

- Determines bolt temperatures and temperature gradients for normal operating conditions

• A structural analysis is to be performed in which failed bolts are inactive

- Stress concentration factors are calculated to determine the peak stresses at the bolt head-to-shank fillet region under normal operating conditions

- Analysis is performed for all loads and load combinations required for an ASME evaluation (stress limits for threaded structural fasteners in subsection NG and Appendix F)

- An evaluation ofjoint stability (or, openness) is also to be performed WCAP- 17096-NP December 2009 Revision 2

A-64 Representative rejected FD bolts are to be removed for laboratory testing to confirm the UT inspection results and IGSCC mechanism, or to identify any other failure mechanism(s)

NOTE: One alternative to removing representative UT rejected bolts for laboratory examination is to re-inspect all bolts at the next refueling outage if continued operation for one additional fuel cycle can be supported by a technical evaluation. Other possible options such as replacement of a predefined bolt pattern may also be pursued.

Based on the results of FD bolt UT inspection and laboratory test results, perform an evaluation to assess future FD bolt failure potential. The changes to the peak stress at the bolt head-to-shank fillet region as a result of the identified failures should be included for evaluation of increased susceptibility to SCC.

Incorporate the effect of future FD bolt failure into the operability evaluation and re-inspection requirement The general methodology to be used for acceptance criteria for the locking devices will be development of an NDE inspection standard that contains examples of acceptable and unacceptable locking device visual indications. The acceptance of locking devices is evaluated in two ways: a) observations with "failed" bolts and b) observations with all bolts intact. Observations of damaged locking devices with all bolts intact represent a condition very different from that of locking device damage at a bolt location that is failed. In addition, a damaged or missing welded locking clip versus a crimpled locking cup potentially represents different initiating phenomena that need to be evaluated.

Analytical efforts for the FD bolts could be performed on a generic basis (for all units except TMI-1),

although unit-specific analyses could decrease the conservatism for some units. The NDE inspection standard could also be developed generically.

Existing Documentation:

A generic flow distributor bolt stress analysis (for all units except TMI-1) was developed for the MRP reactor internals project in 2007 (see AREVA NP document 32-9059506).

An NRC-accepted crack growth rate for Alloy A-286 or Alloy X-750 material is not currently available.

However, the PWROG project (PA-MSC-350) has identified some CGR data that is currently available for a feasibility study of a life assessment approach, if desired.

WCAP- 17096-NP December 2009 Revision 2

B-1 APPENDIX B ACCEPTANCE CRITERIA FLOWCHARTS FOR B&W-DESIGNED COMPONENTS INCLUDED IN MRP-227 WCAP- 17096-NP December 2009 Revision 2

B-2 B.1.1 Core Clamping Items Primary (no expansion link)

Physical Measurement for Core Clamping Wear Determination of dfferetial height of top of ple um rib pads torea*or vessel seating surface, with plenum in reactor ves~sel and fuel asse'rblies removed (See MRP-231 Figure 3l)1 Perform Visual (VT-3) new measurement N examinations'at next indicate wear comparedto the _No defined.fequdency" (currently each 10-year ISI) intrfrene it easuddrn prgnliwseby 0" bility Evaluation OP I-Evaluate:tt-e wear* amount .

-Evatuale the iwea r rote

-E&aluate time for wearto Violat the aceptance 1cutenoioni Repea! measurement at, next 10-year ISI, or K disposition with repair or replacement activities' No ptabt Yes required, or for1-orIo e. .  % ---- disposition with repair or fuel cyces?" reptatiement. activi~ties, No Initiate repair or '

"" replacerntn

• .. activities WCAP- 17096-NP December 2009 Revision 2

B-3 B.1.2 Core Support Shield Cast Outlet Nozzles WCAP- 17096-NP December 2009 Revision 2

B-4 B.1.3 Core Support Shield Vent Valve Discs Ev2Iuaw -6nd'fn for:

expanirwin to CRGTSpdber ,-Opirzb~Ity.Eauto

-Srpaiss ara"yIs.

[Loos pt naly5I 50m~co einInofeiyIe

.Go To 7iov Cbart;

'Icir.CHG Spame castings WCAP- 17096-NP December 2009 Revision 2

B-5 B.1.4 Core Support Shield Vent Valve Retaining Rings and Disc Shaft No subsequent inspectioni is required u~nless. ttgred by now inspection resufts No Yes Repeal Inspecon at next Yes ,O-yeat ISI, or' replacement activities WCAP- 17096-NP December 2009 Revision 2

B-6 B.1.5 Upper Core Barrel Bolts and Locking Devices Primary Upper Core Barrel Bolt Ultrasonic (UT) and Locking Device (VT-3) Examinations 100% of accessible bolts (See MRP-231 Figure 3-7)

-UT Indicationrs in bolt No

,Yes Confirm UT Indication and Determine Failure Mechanism

-Remove rep-esentative bolts or locking devices for laboratory examination, (refer to Section Al).

-If needed, use VT-1 for locking devices UT indication No

< confirmed? >

Yes Evaluate conditon for Operability Evaluabon expansion to LCB, UTS,

-Operability evaluation incorporating futur LTS, FD, SSHT, and Shock potential Pad Bolts I

V Repeat inspection at next Is condition acceptable Yes I 0-year ISI, or SNo for 10 or more clisposifion with repair or years? replacement activities NO

ýon Repeat inspection as 1ý Is condition acceptable Yes required, or r 1 or more t. disposition with repair or f uel cycles? replacement activities Inifiaterepairor replacement activities V

Revise inspection plan based on repair/replacem ent strategy WCAP-17096-NP December 2009 Revision 2

B-7 B.1.6 Lower Core Barrel Bolts and Locking Devices Primary Lower Core Barrel Bolt Ultrasonic (UT) and Locking Device (VT-3) Examinations 100% of accessible bolts (See MRP-231 Figure 3-8t)

.Doany relevant conclitionsexist? Repeat inspection at

-UT in ications in bolt No next defined frequency

-Distorted, loose, broken, (currently each 10-year or missing ISl) locking devices

<*Ies Confirm UT Indication and Determine Failure Mechanism

-Remove representative bolts or locking devices for laboratory examination (refer to Section A.1

-Ifneeded, use VT-1 for locking devices Repeat inspection at next 10-year I$1, or disposition with repair or replacement activltles WCAP- 17096-NP December 2009 Revision 2

B-8 B.1.7 Baffle-to-Former Bolts Primary Baffle-to-Former Bolt Ultrasonic (UT) Examinations 100% of accessible bolts (See MRP-231 Figure 3-2)

Do any relevant Repeat inspection at conditions exist? No next defined frequency (currently after 10 to 15

-UT indications additional years) 4 Yes Confirm UT indication and determine failure mechanism

-Remove representative bolts for laboratory examination (refer to Section A.1)..

I LITindications No Confirmed?

Yes I

Evaluate condition for expansion to baffle-to-baffle Operability Evaluation and core barrel-to-former -Minimum number of bolts and pattern required bolts -Operability evaluation incorporating future failure potential Repeat inspection at next Is condition acceptable Yes 10-year ISI, or for 10 or more disposition with repair or years? replacement activities No Repeat inspection as Go To Flow Chart Is condition acceptable Yes required, or for baffle-to-baffle A. disposition with repair or for 1 or more and core barrel-to-fuel cycles? replacement activities former bolts No Initiate repair or SRevise inspection plan based on replacement activities repair/replacement strategy WCAP- 17096-NP December 2009 Revision 2

B-9 B.1.8 Baffle Plates Primary Baffle Plate Visual (VT-3) Examinations 100% of ao*essible surface vhln 14rnch wround each flow holWssl hole and bolt (Se MRP-2l Figum .3-2)

-Do any relevant Repeat inspection at conditions est? No -0next defined frequency

-delectable cracking (currtly each 10-year

-piece(s) "ace or missing ISI)

Yes racteriza Doe. condition e5 require VT-I, ET, or UT or better charate[zton?

NO Perform VT-1, ET, or UT to ch=acrize length or length

'and depth of observation, Yes V

-Evaluate crack grOwth . '

'Evatuale critical cra-k eGae

.Crack oponing disp,1acement (COO) analysis

-Bypass anaiysts - -

-Loose parts analysis

! i No expanson Initiate repair or replacement actites

-ýRevlmnpeclrplan base don repairfreptaceen-t sieIrAlegy WCAP- 17096-NP December 2009 Revision 2

B-10 B.1.9 Locking Devices of Baffle-to-Former and Internal Baffle-to-Baffle Bolts Primary Locking Device for Baffle-to-Former Bolt and Internal Baffle-to-Baffle Bolt Visual (VT-3) Examinations 100% oaaccessible ockig"devices (See MRP.231 Figure 3-2) condiftiareWei bo anyelaa. Repeat Inspection at

-detectablea crading, NO next defined frquency locksingWto dri de~ces. (wu~otty each I O-year deAes Ylockng No Yes Evaluatet Canditcfli~on doai-ptlth to iocidng devies for external baffl-o4o-batflle bolts and axe

, IhT1-l0-10MI tiCmes`ol (ctsrnse if UT inspecivon of baffl-to-,'more bols is needed it,

.... y .aVenot been UT Inspected recentlly) '

N- eepnvio No tare repsekor ____ Revie inspecation "ptan based'.s nnptcacenen activities ntpatrepItacneii areeg WCAP- 17096-NP December 2009 Revision 2

B-11 B.I.1O Guide Block Dowel Welds No Yes V

Evaluate condition for, expansboi to Alloy X-750 - Operabi!4 Evalualkom dowel locking welds in the -Loose parts analysis , .

upper a3nd lower grid fuel -rripact to the function of guide block assembty suppot pads -Delermine number of required guide Moks Repeat inspection at next "Does NO 10-'year l.r condition trigger disposition with repair'or expansion replacementl ac:eis.

Yes 'oepno V

Go To PlowChart for Alloy X-7$0 dowet lockintg welds in the upperand loweorgrid uel assemnb suppo

• pads WCAP- 17096-NP December 2009 Revision 2

B-12 B.1.11 Incore Monitoring Instrumentation Guide Tube Spiders and Welds Yes Evalua to condition for Operabitdy Evaluation expanslor) to CtRGT Spacer -Evaluate minimum number of spider leg needed Castings and LoVer ýuel -Impact to the func66in of the spider, IMI guide tubes. and Assembly Support Pad Items tncore instrumentation

-Loose.parts analysis Go To Flow Chart for CRGT Spacer Casli*s I and Lowr FuIel Assembly Support Pad Itemqs WCAP- 17096-NP December 2009 Revision 2

B-13 B.2 EXPANSION COMPONENT ITEMS Logic charts for each of the Expansion component items are provided in this section. A separate sub-section is provided for each component item logic chart.

WCAP- 17096-NP December 2009 Revision 2

B- 14 B.2.1 Upper and Lower Grid Fuel Assembly Support Pad Dowel Welds WCAP- 17096-NP December 2009 Revision 2

B-15 B.2.2 Control Rod Guide Tube Spacer Castings Inliataerepair orlvE

epQarPI~bqfiqtA strategy WCAP- 17096-NP December 2009 Revision 2

B- 16 B.2.3 Upper Thermal Shield Bolts and Locking Devices No subsequent inspection is requi'ed unMess trggered by new inspection results WCAP- 17096-NP December 2009 Revision 2

B-17 B.2.4 Surveillance Specimen Holder Tube Studs/Nuts or Bolts and Locking Devices Expansion (CR-3 and DB only)

Surveillance Specimen Holder Tube StudlNut or Bolt Ultrasonic (UT) and Locking Device (VT-3) Examinations 100% of accessible bols Doayrlvn No subsequent inspection is required unless triggered by new inspection results codtin eit Confirm UT Indication and Determine Failure Mechanism

-Remove representative bolts or bcking devices for laboratory examination .(refer to Section A.2)

-if needed, use VT-1 for locking devices V

No WCAP- 17096-NP December 2009 Revision 2

B-18 B.2.5 Core Barrel Cylinder and Former Plates WCAP- 17096-NP December 2009 Revision 2

B-19 B.2.6 Baffle-to-Baffle Bolts Core Barrel-to-Former Bolts I Expansion Baffle-to-Baffle Bolt and Core Barrel-to-Former Bolt-No inspection requirement for internal baflie-0-baffie bols. Eternal baffle-ta-baffle bolls'and'cre barrel to-tonner bolftaroe inac i I(1SeeM'RP-231 Flgue. 3-2') .

Ins,peIofaccoIrding to theý Yes Insec Inaccessible.

ne roe bolts by dosassembly or new lisp~~a V

iS con'dltJon epbe for lor more

• INo Initiate repair or inpcin Iiispectlan No subsequent eut is required unless tridgred by new WCAP- 17096-NP December 2009 Revision 2

B-20 B.2.7 Locking Devices for External Baffle-to-Baffle and Core Barrel-to-Former Bolts Expansion Locking Devicefor Baffle-to-Baffle Bolt and Core Barrel-to-Former Bolt EOwnal:baffle-lo-balnlo bolls and cor barrel-oWlormor bolts are Inaccessible (See MRP-231 Rguro 3-2)

Yes of Inaccessible Wtollocations by disassembly No V

Opearablity EvsIuallon

-Loose parts anali.sis

-hImpacllo the baffle-to-batlle bolts and core barteHo-former bolts opemrablity analysts Inoorporallig lultre failure potential WCAP- 17096-NP December 2009 Revision 2

B-21 B.2.8 Lower Grid Fuel Assembly Support Pad Items WCAP- 17096-NP December 2009 Revision 2

B-22 B.2.9 Lower Grid Shock Pad Bolts and Locking Devices Expansion (TMI-1 only)

Shock Pad Bolt Ultrasonic (UT) and Locking Device (VT-3) Examinations 100% of accessible bolts (See MRP-231 Figure 3-4) codtoseit Nossubsequent "

l a examnat inspection results ViYes Mneeed.use T-iforlocking devices Confirm UT endication and Determine Failure Mechanism

-Remove representative bolts or locking devices for laboratory examination (refer to Section A.2).

-Ifyneededmuse VT-1 for locking devices I

Evaluation eOperability failure

-- Operability evaluation incorporating future potent~

IrT indctonN

~Repeat inspection at next Yes 10-yer ISI, or for 10oracceptable S

Is condition more disposition with repair or f years? replacement activities, No Repeat inspection as; S nditioae for re required,

.......... di spo sition or with repaireor ereplacement activities iNo Initit re aro

' ".replacementln~i r*P;0activities I, V

Revise inspection plan based on2 repairfreplacement strategy WCAP- 17096-NP December 2009 Revision 2

B-23 B.2.10 Lower Thermal Shield Bolts and Locking Devices Expansion Lower Thermal Shield StudlNut or Bolt Ultrasonic (UT) and Locking Device (VT-3) Examinations 100% of accessible bolts (See MRP-231 Figure 3-8)

Do any relevart conditions. exist? No subsequent 4JT indications in bolt No . inspection is reqiired

-Distorted, loose, broken, urness triggered by new or missing inspecton results locking devices

<*Yffi as >

Confirm UT Indication and Determine Failure Mecharism laboratory examination (refer to Section A2)

-If needed, use VT-1 for locking devices I

YV Operability Evaluation

-Operability evaluation incorporating future failure potential .

Repeat inspection at next Isccndifion acceptable Yes 10-year ISI, or for l~or more disposition with repair or years? replacement activities No Repeat inspection as Ismondition acceptable Yes I required, or, s en1ormore, a- .. disposition with repair or furelcycles?- replacement activities No Initiate repair or replacement activities V

Revise inspecon plan based on repal/replacement strategy WCAP- 17096-NP December 2009 Revision 2

B-24 B.2.11 Flow Distributor Bolts and Locking Devices Expansion Flow Distributor Bolt Ultrasonic (UT) and Locking Device (VT-3) Examinations 100% of accessible bolts (See MRP-231 Figure 3-8)

Do any relevant conditions exist? No subsequent

-UT indicaflons, In bok No I. inspection is required

-Distorted, loose, broken. unless triggered by or missing primary ierns again locking devices WCAP- 17096-NP December 2009 Revision 2

C-1 APPENDIX C ACCEPTANCE CRITERIA METHODOLOGY AND DATA REQUIREMENTS FOR COMBUSTION ENGINEERING COMPONENTS INCLUDED IN MRP-227 CE Primary and Expansion Components CE-ID: 1 Core Shroud Assembly (Bolted) - Core Shroud Bolts CE-ID: 1.1 Core Shroud Assembly (Bolted) - Barrel-Shroud Bolts CE-ID: 1.2 Core Shroud Assembly (Bolted) - Core Support Column Bolts CE-ID: 2 Core Shroud Assembly (Welded) - Welds CE-ID: 2.1 Core Shroud Assembly (Welded) - Remaining Axial Welds CE-ID: 3 Core Shroud Assembly (Welded - Full Height) - Shroud Plates CE-ID: 3.1 Core Shroud Assembly (Welded - Full Height) - Axial Welds, Ribs and Rings CE-ID: 4 Core Shroud Assembly (Bolted) - Assembly CE-ID: 5 Core Shroud Assembly (Welded) - Assembly CE-ID: 6 Core Support Barrel Assembly - Upper (Core Support Barrel) Flange Weld CE-ID: 6.1 Core Support Barrel Assembly - Lower Core Barrel Flange CE-ID: 6.2 Core Support Barrel Assembly - Remaining Core Barrel Assembly Welds CE-ID: 6.3 Lower Support Structure - Core Support Column Welds CE-ID: 7 Core Support Barrel Assembly - Lower Flange Weld CE-ID: 8 Lower Support Structure - Core Support Plate CE-ID: 9 Upper Internals Assembly - Fuel Alignment Plate CE-ID: 10 Control Element Assembly - Instrument Guide Tubes CE-ID: 10.1 Control Element Assembly - Remaining Instrument Guide Tubes CE-ID: 11 Lower Support Structure - Deep Beams WCAP- 17096-NP December 2009 Revision 2

C-2 CE-ID: 1 Core Shroud Assembly (Bolted)

Core Shroud Bolts Category: Primary Applicability: Bolted plant designs Degradation Effect: Cracking (IASCC, fatigue)

Expansion Link: Core support column bolts, barrel-shroud bolts Function: The shroud-former bolts fasten the shroud plates to the barrel-former structure.

Inspection Method: Baseline volumetric (UT) examination between 25 and 35 EFPY, with subsequent examination after 10 to 15 additional EFPY to confirm stability of bolting pattern.

Re-examination for high-leakage core designs requires continuing inspections on a 10-year interval.

Coverage: 100% of accessible bolts, or as supported by plant-specific justification. Heads are accessible from the core side. UT accessibility may be affected by complexity of head and locking device designs.

See MRP-227 Figure 4-24.

Observable Effect: UT should reliably detect flaws greater than 30% through-shaft cracking.

Failure Failure Mechanism: Known IASCC cracking of similar highly irradiated bolts has been reported.

Failure Effect: Loss of structural stability Failure Criteria: Require a minimum bolting pattern Methodology Goal: Must demonstrate that projected number of additional bolt failures will not threaten minimum pattern prior to next scheduled inspection.

Data Requirements: Loads Bolting patterns Shroud design Fast neutron (dpa) distribution in core shroud Projected bolt failure rate Minimum bolting pattern analysis WCAP- 17096-NP December 2009 Revision 2

C-3 CE-ID: 1 Core Shroud Assembly (Bolted)

Core Shroud Bolts Analysis: The observed pattern of failed bolts must meet the pre-defined acceptable bolt pattern and have a reasonable margin to protect against additional failures during the inspection interval. The margin is defined in terms of the number of intact bolts beyond the number required for the minimum bolting pattern. The margin, M, at any time is simply:

M = N - Nreq - Nf where N = total number of shroud-former bolts Nreq = number of shroud-former bolts in minimum acceptable pattern Nf = number of failed bolts.

Assuming that there are no failed bolts at the beginning of life, the initial margin is simply: (N - Nreq). For operation through the next 10-15 EFPY interval, require that no more than 50% of initial margin be consumed at the time of the first inspection.

Acceptance Criteria: Procedures for establishing acceptable bolting patterns or the baffle-to-former bolts in Westinghouse-designed plants have been established in [13]. This methodology has been reviewed and accepted by the NRC in a Safety Evaluation in 1998 (TAC No. MA 1152).

The same methodology should be applied to the two operating CE plants with bolted core shrouds.

1. Observed pattern of unfailed bolts meets pre-defined acceptance criteria.

Approach: No generic effort required. Only two plants are affected December 2009 17096-NP WCAP- 17096-NP December 2009 Revision 2

C-4 CE-ID: 1.1 Core Shroud Assembly (Bolted)

Barrel-shroud Bolts Category: Expansion Applicability: Bolted plant designs Degradation Effect: Cracking (IASCC, fatigue)

Expansion Link: Core shroud bolts Function: Maintain structural integrity of barrel-shroud structure.

Inspection Method: Volumetric (UT) examination, with initial and subsequent examination frequencies dependent on the results of core shroud bolt examinations.

Coverage: 100% (or as supported by plant-specific justification) of barrel-shroud and guide lug insert bolts with neutron fluence exposures > 3 displacements per atom (dpa).

Observable Effect UT should reliably detect flaws greater than 30% through-shaft cracking.

Failure Failure Mechanism: Cracking by combined effects of IASCC and fatigue.

Failure Effect: Inability to maintain structural stability Failure Criteria: Require a minimum bolting pattern.

Methodology Goal: Must demonstrate a minimum bolting pattern.

Data Requirements: Bolting patterns Shroud design Fast neutron (dpa) distribution in core shroud Projected bolt failure rate Minimum bolting pattern analysis Analysis: The observed pattern of failed bolts must meet the pre-defined acceptable bolt pattern and have a reasonable margin to protect against additional failures during the inspection interval. The margin is defined in terms of the number of intact bolts beyond the number required for the minimum bolting pattern. The margin, M, at any time is simply:

M = N - Nreq - Nf where N = total number of barrel-former bolts Nreq = number of barrel-former bolts in minimum acceptable pattern Nf = number of failed bolts.

Assuming that there are no failed bolts at the beginning of life, the initial margin is simply: (N-Nreq). For operation through the next 10-15 EFPY interval, require that no more than 50% of initial margin be consumed at the time of the first inspection.

WCAP- I7096-NP December 2009 Revision 2

C-5 CE-ID: 1.1 Core Shroud Assembly (Bolted)

Barrel-shroud Bolts Acceptance Criteria: Procedures for establishing acceptable bolting patterns or the barrel-to-former bolts in Westinghouse designed plants have been established in [13]. This methodology has been reviewed and accepted by the NRC in a Safety Evaluation in 1998 (TAC No. MAI 152).

The same methodology should be applied to the two operating CE plants with bolted core shrouds.

1. Observed pattern of unfailed bolts meets pre-defined acceptance criteria.

Approach: No generic effort required. Only two plants are affected WCAP- 17096-NP December 2009 Revision 2

C-6 CE-ID: 1.2 Core Shroud Assembly (Bolted)

Core support column bolts Category: Expansion Applicability: Bolted plant designs Degradation Effect: Cracking (IASCC, fatigue)

Expansion Link: Core shroud bolts Function: Attach core support columns to core support plate.

Inspection Method: Ultrasonic (UT) examination, with initial and subsequent examination frequencies dependent on the results of core shroud bolt examinations.

Coverage: 100% (or as supported by plant-specific analysis) of core support column bolts with neutron fluence exposures > 3 dpa.

Observable Effect: UT should reliably detect flaws greater than 30% through-shaft cracking.

Failure Failure Mechanism: IASCC and fatigue Failure Effect: Loss of structural stability Failure Criteria: Determine minimum number of support columns required to maintain structural integrity.

Methodology Goal: Establish functional requirements for core support columns.

• During normal operation system of support columns should resist core plate deformation due to mechanical or thermal loading. Core plate requirements for "flatness" and fuel assembly alignment.

" During limiting accident transient system must maintain structural integrity.

Data Requirements: Loads on core support plate.

Displacement tolerances on lower core plate.

Analysis: See MRP-227 Figures 4-16 and 4-33. Build FEA model of lower support structure that includes support columns and core support plate. Model should be capable of removing individual column or breaking attachment to lower core support plate. Would require multiple iterations to establish "minimum acceptable patterns" of core support columns and support column bolts.

Structural model must be run for functional requirements A and B.

Determine margin for additional failures.

1. Assume number of failures in next 10 years is equal to number observed to date.

N = # of support columns Nf= # of observed flawed columns Nreq = # of columns in relevant minimum pattern Margin = N-Nreq WCAP- 17096-NP December 2009 Revision 2

C-7 CE-ID: 1.2 Core Shroud Assembly (Bolted)

Core support column bolts Acceptance Criteria: Require that no more of 1/2 of columns in margin are failed:

Nf < (N-Nreq)/2 Approach: Generic program to share first-of-a-kind effort. (See W-ID: 2-1)

• Pilot analysis of lower support structure to identify critical issues.

• Expect final acceptance based on plant-specific analysis.

WCAP- 17096-NP December 2009 Revision 2

C-8 CE-ID: 2 Core Shroud Assembly (Welded)

Category: Primary Applicability: Plant designs with core shrouds assembled in two vertical sections Degradation Effect: Cracking (IASCC)

Expansion Link: Remaining axial welds Function: 1. Maintain core geometry.

2. Direct coolant flow.

Inspection Method: Enhanced visual (EVT-1) examination no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

Coverage: Axial and horizontal weld seams at the core shroud re-entrant comers as visible from the core side of the shroud, within six inches of central flange and horizontal stiffeners.

See MRP-227 Figures 4-12 and 4-14.

Observable Effect: Cracking Failure Failure Mechanism: SCC Failure Effect: 1. Core damage caused by event. Require maintenance of coolable geometry.

2. Damage to peripheral fuel assemblies.

3. Through-wall crack provides leak path through shroud.

Failure Criteria: Observed flaw will not grow to critical crack size for crack initiation during limiting transient event prior to next planned inspection.

No observable damage in corresponding sections of peripheral fuel assemblies.

Methodology Goal: Perform flaw-tolerance analysis to demonstrate that crack will not grow to exceed crack initiation size limit during limiting transient events.

Data Requirements: 1. Normal operating loads (plant specific)

2. Elastic-plastic K solution for normal operation (geometry dependent)

3. Fast neutron fluence (or dpa) at crack location (plant specific)

4. IASCC crack growth rate curve

5. Limiting transient loads (plant specific)

6. K solution for limiting transient (geometry dependent)

7. Irradiated fracture toughness (K,,)

WCAP- 17096-NP December 2009 Revision 2

C-9 CE-ID: 2 Core Shroud Assembly (Welded)

Analysis: 1. Assume through-wall crack of length (L) corresponding to visual indication.

2. Estimate normal operating loads at crack tip as determined by:

• Weld residual stresses

• Irradiation induced stress relaxation

• Swelling induced stresses

• Normal operation (Delta-P, Delta-T, flow, dead weight)

Note: May be reasonable to assume that residual stress and stress relaxation are offsetting factors.

3. Obtain stress intensity factor (K) solution corresponding to crack at comer with described loads.

4. Construct models for fatigue and IASCC crack growth rates as a function of K.

5. Integrate crack growth rate over next inspection interval to estimate crack length.

Note: This may be accomplished numerically.

6. Estimate limiting transient load (presumably LOCA).

7. Obtain stress intensity factor (Kapp) solution corresponding to crack at comer with transient loads.

8. For center of core shroud location, use limiting fracture toughness, Kjc, for highly irradiated material.

Acceptance Criteria: Structure is acceptable if Kapp < Kic Approach: Expect calculation to be plant specific

• Define general load conditions at weld seams.

• Define K-solution for loading at weld seams.

WCAP- 17096-NP December 2009 Revision 2

C-i10 CE-ID: 2.1 Core Shroud Assembly (Welded)

Remaining Axial Welds Category: Expansion Applicability: Plant designs with core shrouds assembled in two vertical sections Degradation Effect: Cracking (IASCC)

Expansion Link: Core shroud plate-former plate weld.

Function: 1. Maintain core geometry.

2. Direct coolant flow.

Inspection Method: Enhanced visual (EVT-1) examination, with initial and subsequent examination frequencies dependent on the results of the core shroud weld examinations.

Coverage: Axial weld seams other than the core shroud re-entrant comer welds at the core mid-plane.

Observable Effect: Cracking Failure Failure Mechanism: IASCC Failure Effect: 1. Core damage caused by event - require maintenance of coolable geometry.

2. Damage to peripheral fuel assemblies.

3. Through-wall crack provides leak path through shroud.

Failure Criteria: 1. Observed flaw will not grow to critical crack size for crack initiation during limiting transient event prior to next planned inspection.

2. No observable damage in corresponding sections of peripheral fuel assemblies Methodology Goal: Demonstrate that cracks in axial welds are stable.

Data Requirements: 1. Normal operating loads (plant specific)

2. Elastic-plastic K solution for normal operation (geometry dependent)

3. Fast neutron fluence (or dpa) at crack location (plant specific)

4. lASCC crack growth rate curve

5. Limiting transient (potentially LOCA) loads (plant specific)

6. K solution for limiting transient (geometry dependent)

7. Irradiated fracture toughness (Kic)

WCAP- 17096-NP December 2009 Revision 2

C-11 CE-ID: 2.1 Core Shroud Assembly (Welded)

Remaining Axial Welds Analysis: 1. Assume through-wall crack of length (L) corresponding to visual indication.

2. Estimate normal operating loads at crack tip as determined by:

• Weld residual stresses

• Irradiation induced stress relaxation

• Swelling induced stresses

• Normal operation (Delta-P, Delta-T, flow, dead weight)

Note: May be reasonable to assume that residual stress and stress relaxation are offsetting factors.

3. Obtain stress intensity factor (K) solution corresponding to crack at comer with described loads.

4. Construct models for fatigue and IASCC crack growth rates as a function of K.

5. Integrate crack growth rate over next inspection interval to estimate crack length.

Note: This may be accomplished numerically.

6. Estimate limiting transient load (presumably LOCA).

7. Obtain stress intensity factor (Kapp) solution corresponding to crack at comer with transient loads.

8. For center of core shroud location, use limiting fracture toughness, Kjc, for highly irradiated material.

Acceptance Criteria: Structure is acceptable if Kapp < Kic Approach: Plant-specific analysis.

• Require flaw tolerance handbook/methodology based on flaw location and direction.

WCAP- 17096-NP December 2009 Revision 2

C- 12 CE-ID: 3 Core Shroud Assembly (Welded)

Shroud Plates Category: Primary Applicability: Plant designs with core shrouds assembled with full-height shroud plates Degradation Effect: Cracking (IASCC)

Expansion Link: Remaining axial welds, ribs and rings Function: 1. Maintain core geometry.

2. Direct coolant flow.

Inspection Method: Enhanced visual (EVT-1) examination no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

Coverage: Axial weld seams at the core shroud re-entrant comers, at the core mid-plane (+/-three feet in height) as visible from the core side of the shroud. See MRP-227, Figure 4-13.

Observable Effect: Cracking Failure Failure Mechanism: IASCC Failure Effect: 1. Core damage caused by event - require maintenance of coolable geometry.

2. Damage to peripheral fuel assemblies.

3. Through-wall crack provides leak path through shroud.

Failure Criteria: 1. Observed flaw will not grow to critical crack size for crack initiation during transient loading condition event prior to next planned inspection.

2. No observable damage in corresponding sections of peripheral fuel assemblies.

Methodology Goal: Perform flaw-tolerance analysis to demonstrate that crack will not grow to exceed crack initiation size limit during limiting transient events.

Data Requirements: 1. Normal operating loads (plant specific)

2. Elastic-plastic K solution for normal operation (geometry dependent)

3. Fast neutron fluence (or dpa) at crack location (plant specific)

4. IASCC crack growth rate curve

5. Limiting transient loads (plant specific)

6. K solution for limiting transient (geometry dependent)

7. Irradiated fracture toughness (Kic)

WCAP- 17096-NP December 2009 Revision 2

C-13 CE-ID: 3 Core Shroud Assembly (Welded)

Shroud Plates Analysis: 1. Assume through-wall crack of length (L) corresponding to visual indication.

2. Estimate normal operating loads at crack tip as determined by:

• Weld residual stresses

• Irradiation induced stress relaxation

• Swelling induced stresses.

• Normal operation (Delta-P, Delta-T, flow, dead weight)

Note: May be reasonable to assume that residual stress and stress relaxation are offsetting factors.

3. Obtain stress intensity factor (K) solution corresponding to crack at comer with described loads.

4. Construct models for fatigue and IASCC crack growth rates as a function of K.

5. Integrate crack growth rate over next inspection interval to estimate crack length.

Note: This may be accomplished numerically.

6. Estimate limiting transient load (presumably LOCA).

7. Obtain stress intensity factor (Kapp) solution corresponding to crack at comer with transient loads.

8. For center of core shroud location, use limiting fracture toughness, K1 c, for highly irradiated material.

Acceptance Criteria: Structure is acceptable if Kapp < Ki.

Approach: No generic analysis: Only one utility with this design.

WCAP- 17096-NP December 2009 Revision 2

C-i14 CE-ID: 3.1 Core Shroud Assembly (Welded)

Remaining Axial Welds, Ribs and Rings Category: Expansion Applicability: Plant designs with core shrouds assembled with full-height shroud plates Degradation Effect: Cracking (IASCC)

Expansion Link: Shroud plates of welded core shroud assemblies Function: I. Maintain dimensional stability of core shroud plus ribs and rings.

Inspection Method: Enhanced visual (EVT-1) examination, with initial and subsequent examination frequencies dependent on the results of the core shroud weld examinations.

Coverage: Axial weld seams other than the core shroud re-entrant comer welds at the core mid-plane.

Observable Effect: Cracking Failure Failure Mechanism: IASCC Failure Effect: 1. Deformation of core barrel leads to interaction with fuel.

2. Unable to withstand limiting transient loads due to a loss of structural support.

3. Generation of loose parts.

Failure Criteria: Welds with observable cracks assumed failed.

Components with cracks in all attachment welds considered as potential loose part.

Require minimum acceptable support structure to withstand limiting transient forces.

Methodology Goal: Demonstrate that:

1. Cracks in axial welds are stable

2. Cracks in ribs and rings will not generate loose parts Data Requirements: 1. Normal operating loads (plant specific)

2. Elastic-plastic K solution for normal operation (geometry dependent)

3. Fast neutron fluence (or dpa) at crack location (plant specific)

4. IASCC crack growth rate curve

5. Limiting transient loads (plant specific)

6. K solution for limiting transient (geometry dependent)

7. Irradiated fracture toughness (K1 c)

Analysis: For remaining axial welds:

1. Assume through-wall crack of length (L) corresponding to visual indication.

WCAP- 17096-NP December 2009 Revision 2

C-15 CE-ID: 3.1 Core Shroud Assembly (Welded)

Remaining Axial Welds, Ribs and Rings

2. Estimate normal operating loads at crack tip as determined by:

• Weld residual stresses

• Irradiation induced stress relaxation

• Swelling induced stresses

• Normal operation (Delta-P, Delta-T, flow, dead weight)

Note: May be reasonable to assume that residual stress and stress relaxation are offsetting factors.

3. Obtain stress intensity factor (K) solution corresponding to crack at comer with described loads.

4. Construct models for fatigue and IASCC crack growth rates as a function of K.

5. Integrate crack growth rate over next inspection interval to estimate crack length.

Note: This may be accomplished numerically.

6. Estimate limiting transient load (presumably LOCA).

7. Obtain stress intensity factor (Kapp) solution corresponding to crack at comer with transient loads.

8. For center of core shroud location, use limiting fracture toughness, K1,, for highly irradiated material.

9. Structure is acceptable if Kapp < KIc.

For ribs and rings:

Prepare for examination by conducting a failure modes and effects analysis (FMEA) to identify full range of potential relevant observations prior to inspection. Primary concern:

• Fracture of weld between shroud and ring

• Fracture of welds in stiffeners The major effects of these failure mechanisms are expected to be:

• Loss of stability in shroud structure (possible deformation and interaction with fuel assembly)

• Loose parts Acceptance Criteria: For axial welds: Kapp < Kic A plant-specific plan should be developed for evaluating and mitigating the potential relevant conditions related to weld failures in ribs, rings or stiffeners. The evaluation should consider any previously reported observations.

Approach: No generic analysis: Only one utility with this design.

WCAP- 17096-NP December 2009 Revision 2

C-16 CE-ID: 4 Core Shroud Assembly (Bolted)

Assembly Category: Primary Applicability: Bolted plan designs Degradation Effect: Distortion (Void Swelling)

Expansion Link: None Function: Provide support, guidance, and protection for the reactor core.

Provide a passageway for the distribution of the reactor coolant flow to the reactor core.

Provide gamma and neutron shielding for the reactor vessel.

Inspection Method: Visual (VT-3) examination no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examinations on a 10-year interval..

Coverage: Core side surfaces as indicated. See Figures 4-25 and 4-26 of MRP-227.

Observable Effect: Degradation of general condition as described above.

Failure Failure Mechanism: Void swelling Failure Effect: 1. Interference with fuel assemblies

2. Obstruction of coolant flow

3. Loose parts generation

4. Distortion/misalignment of core

5. Local temperature peaks

6. Degradation of control rod insertability

7. Baffle jetting Failure Criteria: No relevant observations Methodology Goal: A plant-specific plan should be developed for evaluating and mitigating the potential relevant conditions. The evaluation should consider any previously reported observations.

Data Requirements: Baseline data on previous visual examinations of core shroud.

Performance records for peripheral fuel assemblies.

Loose parts monitoring data.

Analysis: Prepare for examination by conducting a failure modes and effects analysis (FMEA) to identify full range of potential relevant observations prior to inspection. Failure mechanisms considered should include:

• Broken or missing locking devices,

• Protruding bolt heads Missing bolts or bolt heads.

WCAP- I7096-NP December 2009 Revision 2

C-17 CE-ID: 4 Core Shroud Assembly (Bolted)

Assembly

• Distortion or displacement of shroud plates

• Gross cracking of shroud plates

• Gaps at plate joints

• Interaction with fuel assemblies Acceptance Criteria: Determined by FMEA Approach: FMEA should address plant-specific practices and priorities. Some generic work possible to outline issues and options to be addressed in FMEA.

WCAP- 17096-NP December 2009 Revision 2

c-i18 CE-ID: 5 Core Shroud Assembly (Welded)

Assembly Category: Primary Applicability: Plant designs with core shrouds assembled in two vertical sections Degradation Effect: Distortion (void swelling), as evidenced by separation between the upper and lower flanges.

Expansion Link: None Function: Provide support, guidance, and protection for the reactor core.

Provide a passageway for the distribution of the reactor coolant flow to the reactor core.

Provide gamma and neutron shielding for the reactor vessel.

Inspection Method: Visual (VT-1) examination no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examinations on a 10-year interval.

Coverage: If a gap exists, make three to five measurements of gap opening from the core side at the core shroud re-entrant comers. Then, evaluate the swelling on a plant-specific basis to determine frequency and method for additional examinations. See MRP-227 Figures 4-12 and 4-14.

Observable Effect: Seam between upper and lower sections should appear even and consistent with any historical records.

Evidence of gaping between plates at protruding comers should be considered a relevant condition.

Failure Failure Mechanism: Void Swelling Failure Effect: 1. Potential leakage through shroud.

2. Significant distortion may interfere with peripheral fuel assemblies.

3. Condition is a precursor to high stresses and potential cracking at weld seams.

Failure Criteria: 1. Damage on corresponding peripheral fuel assemblies.

2. Gap size implies peak shroud swelling > 5% by volume.

Methodology Goal: A plant-specific plan should be developed for evaluating and mitigating the potential relevant conditions. The evaluation should consider any previously reported observations.

Data Requirements: 1. Gap size

2. Swelling deformation model of shroud

3. Shroud fluence distribution

4. Shroud temperature distribution WCAP- 17096-NP December 2009 Revision 2

C-19 CE-ID: 5 Core Shroud Assembly (Welded)

Assembly Analysis: Prepare for examination by conducting a failure modes and effects analysis (FMEA) to identify full range of potential relevant observations prior to inspection. Failure mechanisms considered should include:

• Broken or missing locking devices

• Protruding bolt heads

• Missing bolts or bolt heads

• Distortion or displacement of shroud plates

• Gross cracking of shroud plates

• Gaps at plate joints

• Interaction with fuel assemblies Acceptance Criteria: Quantitative evaluation of swelling would require a time dependent structural model that incorporates the effects of void swelling. Temperature gradients caused by gamma heating must be accurately estimated to provide reliable swelling estimates. This detailed evaluation is only required if repeated observations indicate gap is actively growing.

Determined by FMEA Approach: Generic efforts to support inspection.

• Extension of MRP model to look at relationship between swelling and deformation at seam.

• Guideline for issues to be addressed in plant-specific FMEA.

WCAP- 17096-NP December 2009 Revision 2

C-20 CE-ID: 6 Core Support Barrel Assembly Upper (Core Support Barrel) Flange Weld Category: Primary Applicability: All plants Degradation Effect: Cracking (SCC)

Expansion Link: Remaining core barrel assembly welds, core support column welds Function: Primary core support structure.

Inspection Method: Enhanced visual (EVT-1) examination no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examinations on a 10-year interval.

Coverage: 100% of the accessible surfaces of the upper flange weld. See MRP-227 Figure 4-15.

Observable Effect: Stress corrosion cracking Failure Failure Mechanism: SCC Failure Effect: Loss of core support Failure Criteria: Actively growing through-wall flaws require mitigation. Require demonstration that flaw growth is arrested or limited to surface.

An existing through-wall flaw may be acceptable if condition and shape indicate that it is non-growing fabrication flaw.

Methodology Goal: Due to the high fracture toughness of unirradiated stainless steel, the core barrel is a highly flaw tolerant structure and flaw sizes are expected to be very large. However, the core barrel is a critical support structure. Flaw growth in this component is outside the range of normal expectations. Therefore, it has been assumed that the presence of any actively growing through-wall crack would require repair or other mitigation. The goal of the calculation is to demonstrate the crack is stable or not likely to grow through wall.

Data Requirements: 1. Operating loads

2. K Solutions for range of expected crack shapes (lengths and depths)

3. SCC crack growth rate curves

4. Fatigue crack growth rate curve (as backup)

Analysis: Strategy similar to Westinghouse core barrel upper flange weld.

Option 1. Observation on OD of core support barrel Step 1. Determine stress distribution through core support barrel thickness for normal operating conditions (expect peak stress at vessel OD).

Step 2. Obtain stress intensity factor solution for part-through-wall crack as function of surface length (L) and depth (a).

Step 3. Short cracks will be constrained by the stress distribution in the barrel wall. Define the maximum constrained crack length as Lc.

WCAP- 17096-NP December 2009 Revision 2

C-21 CE-ID: 6 Core Support Barrel Assembly Upper (Core Support Barrel) Flange Weld Step 4. OD crack observation is acceptable if L<Lc.

Step 5. IfL > Lc, then must perform UT to determine crack depth (a).

Step 6. Crack is acceptable if K corresponding to a and Lc is less than 20 ksi-in^1/2.

Step 7. All remaining cracks require specific flaw-tolerance analysis.

Option 2. Observation of flaw on ID of core support barrel Step 1. If flaw on ID is smaller than the length (Lc) defined in Option. 1, visually examine the OD surface corresponding to the ID flaw to determine if it is OD-initiated. Crack is acceptable if not through-wall.

Step 2. For a through-wall flaw, apply the OD flaw acceptance criteria from Option 1.

Step 3. All remaining cracks require a geometry-specific flaw-tolerance analysis.

Option 3. Observation of crack on ID of core support barrel Step 1. If flaw on ID is smaller than the length (Lc) defined in Option 1, perform UT exam to determine if the crack is through-wall. Crack is acceptable if not through-wall.

Step 2. For a through-wall flaw, apply the OD flaw acceptance criteria from Option 1.

Step 3. All remaining cracks require a geometry-specific flaw-tolerance analysis.

Acceptance Criteria: Demonstrate that crack is not actively growing or limited to surface as indicted by analysis.

Approach: Plant-specific analysis WCAP- 17096-NP December 2009 Revision 2

C-22 CE-ID: 6.1 Core Support Barrel Assembly Lower Core Barrel Flange Category: Expansion Applicability: All plants Degradation Effect: Cracking (SCC, Fatigue)

Expansion Link: Upper (core support barrel) flange weld Function: Primary core support.

Inspection Method: Enhanced visual (EVT- 1) examination, with initial and subsequent examinations dependent on the results of the upper (core support barrel) flange weld examinations.

Coverage: 100% of accessible welds and adjacent base metal.

See MRP-227 Figure 4-15.

Observable Effect: Cracks Failure Failure Mechanism: SCC or fatigue cracking in weld or weld heat affected zone.

Failure Effect: Loss of core support Failure Criteria: Potential for growth of a through-wall flaw before next planned inspection.

An existing through flaw may be acceptable if condition and shape indicate that it is a non-growing fabrication flaw.

Methodology Goal: Due to the high fracture toughness of unirradiated stainless steel, the core barrel is a highly flaw tolerant structure and flaw sizes are expected to be very large. However, the core barrel is a critical support structure. Flaw growth in this component is outside the range of normal expectations. Therefore, it has been assumed that the presence of any actively growing through-wall crack would require repair or other mitigation. The goal of the calculation is to demonstrate the crack is stable or not likely to grow through wall.

Data Requirements: 1. Operating loads

2. K Solutions for range of expected crack shapes (lengths and depths)

3. SCC crack growth rate curves

4. Fatigue crack growth rate curve (as backup)

Analysis: 1. Perform FEA to determine stress distribution across weld.

2. Evaluate stress distribution to determine surface with highest probability of crack initiation (highest tensile stress).

WCAP- 17096-NP December 2009 Revision 2

C-23 CE-ID: 6.1 Core Support Barrel Assembly Lower Core Barrel Flange

3. Establish criteria for most likely surface.

> K solution for observed crack length indicates diminishing stress intensity with increasing crack length.

- or -

> UT examination indicates that flaw is limited to initiating surface.

4. Establish criteria for least likely surface.

> Require demonstration that observed flaw was not initiated on opposite surface and grown through wall.

• No visual evidence of cracking on opposite surface.

• Visual observation of opposite surface indicates deep, narrow crack inconsistent with actively growing mechanism.

• UT exam indicates that flaw is limited to initiating surface.

5. Flaws that do not meet criteria of 3 and 4 require additional geometry-specific analysis to estimate rate of crack growth and establish acceptable crack lengths.

Reduced inspection intervals may be required.

Acceptance Criteria: Current crack size is explainable by known crack growth rate laws and limited crack growth is projected.

Approach: Plant-specific analysis.

• Require flaw tolerance handbook/methodology based on flaw location and direction.

• MRP-2 10 may have limited relevance.

December 2009 WCAP- 17096-NP WCAIP- 17096-NP December 2009 Revision 2

C-24 CE-ID: 6.2 Core Support Barrel Assembly Remaining Core Barrel Assembly Welds Category: Expansion Applicability: Core support barrel assembly Degradation Effect: Cracking (SCC)

Expansion Link: Upper (core support barrel) flange weld Function: Primary core support Inspection Method: Enhanced visual (EVT- 1) examination, with initial and subsequent examinations dependent on the results of core barrel assembly upper flange weld examinations.

Coverage: 100% of one side of the accessible weld and adjacent base metal surfaces for the weld with the highest calculated operating stress.

Observable Effect: Cracks Failure Failure Mechanism: Due to the high fracture toughness of unirradiated stainless steel, the core barrel is a highly flaw tolerant structure and critical flaw sizes are expected to be very large.

However, the core barrel is a critical core support structure. Flaw initiation and growth in this component is outside the range of normal expectations. Therefore, it has been assumed that the presence of any actively growing through-wall flaw would require repair or other mitigation.

Failure Effect: Potential loss of core support Failure Criteria: Potential for growth of a through-wall flaw before next planned inspection. An existing through flaw may be acceptable if condition and shape indicate that it is a non-growing fabrication flaw.

Methodology Goal: Demonstrate that observed flaws are not actively growing.

Data Requirements: 1. Fast neutron fluence (dpa)

2. Irradiated fracture toughness

3. Operating loads

4. K Solutions for range of expected crack shapes (lengths and depths)

5. SCC crack growth rate curves

6. Fatigue crack growth rate curve (as backup)

WCAP- 17096-NP December 2009 Revision 2

C-25 CE-ID: 6.2 Core Support Barrel Assembly Remaining Core Barrel Assembly Welds Analysis: 1. Flaws in core support barrel above the shroud section will be evaluated assuming active crack growth mechanisms are SCC and fatigue.

2. Flaws in the beltline region of the core support barrel (shroud section) will be evaluated assuming active growth mechanisms are IASCC and fatigue.

3. A fluence estimate at the flaw location is required for all flaws in the beltline region.

4. Normal operating and fatigue loads will be established for core barrel at this location.

5. Determine stress intensity factors for a through-wall crack.

6. Use appropriate crack growth rate models (SCC or IASCC and fatigue) to estimate crack growth rate.

7. If crack growth rate is consistent with observed flaw size:

• Project flaw size through inspection interval using crack growth rate estimate.

• Determine loads during limiting transient.

• Determine stress intensity factor for through-wall crack of projected length.

• For low fluence region assume KI, = 150 ksi-inA1/2

• For beltline region determine lower bound toughness based on fluence estimate.

• If stress intensity factor during transient is less than fracture toughness, flaw is acceptable.

• If stress intensity factor during transient is greater than fracture toughness, proceed to Step 8.

8. If crack growth rate is too low to explain existence of observed crack or flaw not acceptable by Step 7:

• Determine crack depth.

• If crack depth small compared to barrel thickness (<xx inches), then crack is acceptable.

• If crack depth large compared to barrel thickness, the crack is rapidly growing and a detailed analysis is required.

Acceptance Criteria: Current crack size is explainable by known crack growth rate laws and limited crack growth is projected.

Approach: Plant-specific analysis. (See item CE-ID 6.1)

• Require flaw tolerance handbook/methodology based on flaw location and direction.

• MRP-210 may have limited relevance.

WCAP- 17096-NP December 2009 Revision 2

C-26 CE-ID: 6.3 Lower Support Structure Core Support Column Welds Category: Expansion Applicability: All plants except those with core shrouds assembled with full-height shroud Degradation Effect: Cracking (SCC, IASCC, fatigue) including damaged or fractured material Expansion Link: Upper (core support barrel) flange weld Function: The support columns are a primary core support structure. The columns keep the core support plate from sagging or excess thermal deformation.

Inspection Method: Visual (VT-3) examination, with initial and subsequent examinations based on plant evaluation of SCC susceptibility and demonstration of remaining fatigue life.

Coverage: Examination coverage determined by plant-specific analysis.

See MRP-227 Figures 4-16 and 4-31.

Observable Effect: Fracture Potential for fuel assembly misalignment Failure Failure Mechanism: Cracking Failure Effect: Failure of support columns will allow local deformation of core support plate.

Cracks initiating in welds may lead to fracture or loss of attachment to core support plate.

Failure Criteria: Must establish minimum core support column distributions required to maintain core support plate stability. (Alternative would be to demonstrate that a limited number (5) of failures are generally acceptable.)

Methodology Goal: Establish minimum acceptable pattern of core support columns.

Data Requirements: 1. Design criteria used to determine number and spacing of core supports: lower core support plate loads during normal operating and limiting transient conditions, etc.

2. Loads on lower core plate

3. Fluence accumulated by the core support columns

4. Constitutive model for stainless steel properties as a function of irradiation

5. Displacement tolerances on core support plate

6. Geometry Analysis: 1. Establish functional requirements for core support columns.

A. During normal operation, the system of support columns should resist core plate deformation due to mechanical or thermal loading. Core plate requirements for "flatness" and fuel assembly alignment.

B. During limiting accident transient, the system must maintain structural integrity.

WCAP- 17096-NP December 2009 Revision 2

C-27 CE-ID: 6.3 Lower Support Structure Core Support Column Welds

2. Support column analysis assumptions.

A. Assume any column with a crack in main body to have failed.

B. Assume any.column with a crack in the weld to result in failure of the attachment.

3. Structural model of lower support structure.

Model of lower support structure that includes support columns and lower core plate.

Model should be capable of removing individual column or breaking attachment to lower core plate. Would require multiple iterations to establish "minimum acceptable patterns" of core support columns and support column welds.

4. Structural model must be run for functional requirements A and B.

5. Determine margin for additional failures.

Assume number of failures in next 10 years is equal to number observed to date.

N = # of Support Columns Nf = # of Observed Flawed Columns Nreq = # of columns in relevant minimum pattern Margin = N - Nreq Acceptance Criteria: Require that no more of 1/2 of columns in margin are failed:

Nf < (N - Nreq)/2 Approach: Generic program to share first-of-a-kind effort. (See W-ID: 2.1 and CE-ID: 1.2)

• Pilot analysis of lower support structure to identify critical issues.

• Expect final acceptance based on plant-specific analysis.

WCAP- 17096-NP December 2009 Revision 2

C-28 CE-ID: 7 Core Support Barrel Assembly Lower Flange Weld Category: Primary Applicability: All plants Degradation Effect: Cracking (fatigue)

Expansion Link: None Function: Primary core support structure Inspection Method: If fatigue life cannot be demonstrated by time-limited aging analysis (TLAA), enhanced visual (EVT-1) examination is required no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examination on a 10-year interval.

Coverage: Examination coverage to be defined by plant-specific fatigue analysis. See MR.P-227 Figure 4-15.

Observable Effect: Cracking Failure Failure Mechanism: Fatigue Failure Effect: Loss of core support Failure Criteria: Potential for growth of a through-wall flaw before next planned inspection.

An existing through-wall flaw may be acceptable if condition and shape indicate that it is a non-growing fabrication flaw.

Methodology Goal: Demonstrate that observed flaws are not actively growing.

Data Requirements: 1. Operating loads

2. K solutions for range of expected crack shapes (lengths and depths)

3. Fatigue crack growth rate curves

4. SCC crack growth rate curve (as backup)

Analysis: Inspection of this item is required if sufficient fatigue life cannot be demonstrated by normal time-limited aging analysis (TLAA) procedures. Due to general concerns about SCC in structural welds, the same location has been listed as an expansion inspection that would be triggered by observation of cracking in the upper flange weld.

A general outline of TLAA procedures is provided separately. The TLAA process evaluates potential fatigue crack initiation. As part of that evaluation the stress amplitude and frequency must be estimated. If the TLAA indicates that crack initiation is possible, inspection of the indicated locations is required. Fatigue crack growth rates used in establishing acceptance criteria for the inspections should be based on the stress amplitudes and frequencies used in the TLAA.

WCAP- 17096-NP December 2009 Revision 2

C-29 CE-ID: 7 Core Support Barrel Assembly Lower Flange Weld The following analysis parallels the requirements for the expansion inspections.

1. Perform FEA to determine stress distribution across weld.

2. Evaluate stress distribution to determine surface with highest probability of crack initiation (highest tensile stress).

3. Establish criteria for the highest probability surface.

• Demonstrate that a 1/4 thickness flaw of observed length will not grow through barrel wall in planned inspection interval.

4. Establish criteria for the lowest probability surface.

• Require demonstration that observed flaw was not initiated on opposite surface and grown through wall.

- No visual evidence of cracking on opposite surface.

- Visual observation of opposite surface indicates deep, narrow crack inconsistent with actively growing mechanism.

- UT exam indicates that flaw is limited to initiating surface.

5. Flaws that do not meet criteria of Items 3 and 4 require additional geometry-specific analysis to estimate rate of crack growth andestablish acceptable crack lengths.

Reduced inspection intervals may be required.

Acceptance Criteria: Acceptance criteria for TLAA related items are beyond scope of current project.

Approach: TLAA (plant specific)

Potential flaw analysis if inspection required.

• Require flaw tolerance handbook/methodology based on flaw location and direction.

• MRP-210 may have limited relevance.

WCAP- 17096-NP December 2009 Revision 2

C-30 CE-ID: 8 Lower Support Structure Core Support Plate Category: Primary Applicability: All plants with a core support plate Degradation-Effect: Cracking (fatigue)

Expansion Link: None Function: Primary core support. Provides alignment of fuel assembly.

Inspection Method: If fatigue life cannot be demonstrated by time-limited aging analysis (TLAA), enhanced visual (EVT- 1) examination is required no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examination on a 10-year interval.

Coverage: Examination coverage to be defined by plant-specific fatigue analysis. See MRP-227 Figure 4-16.

Observable Effect: Cracking Failure Failure Mechanism: Fatigue Failure Effect: 1. Loss of core support

2. Difficulty in loading fuel due to misalignment Failure Criteria: Displacement of core support plate Methodology Goal: Cracks in the core support plate are expected to grow from hole-to-hole within the plate.

A network of connected cracks is required to allow significant displacement in the plate.

The goal is to demonstrate that the cracking present does not cause enough displacement to affect fuel loading or core support.

Data Requirements: 1. Operating loads

2. K Solutions for range of expected crack shapes (lengths and depths)

3. Fatigue crack growth rate curves

4. IASCC crack growth rate curve and fluence (as backup)

Analysis: Inspection of this item is only required if sufficient fatigue life cannot be demonstrated by normal time-limited aging analysis (TLAA) procedures.

A general outline of TLAA procedures is provided separately. The TLAA process evaluates potential fatigue crack initiation. As part of that evaluation the stress amplitude and frequency must be estimated. If the TLAA indicates that crack initiation is possible, inspection of the indicated locations is required. Fatigue crack growth rates used in establishing acceptance criteria for the inspections should be based on the stress amplitudes and frequencies used in the TLAA.

WCAP- 17096-NP December 2009 Revision 2

C-31 CE-ID: 8 Lower Support Structure Core Support Plate Process steps for establishing allowable crack length in the core support plate:

1. Establish functional requirements for core support plate.

A. During normal operation, the system of support columns should resist core support plate deformation caused by mechanical or thermal loading. The core support plate would have requirements for "flatness" and fuel assembly alignment.

B. During limiting accident transient system must maintain structural integrity.

2. Core support plate analysis assumptions.

A. Assume crack initiates at the hole or holes in plate with highest surface tensile stress.

B. Assume crack propagates to the adjacent hole with highest stress.

3. Structural model of lower support structure.

Model of lower support structure that includes support columns and lower core support plate. Model should be capable of modeling a crack connecting holes in plate (crack tip modeling not required). Evaluate displacement on the surface of the core support plate.

4. Any single observed crack is acceptable if displacement across crack in FEA model meets design requirements for plate.

5. If unable to demonstrate acceptability of a single crack, require detailed flaw analysis.

6. Optional determination of margin for additional cracking. Repeat evaluation for multiple cracks connecting adjacent holes. Determine number and pattern of connected holes to violate design requirements.

Acceptance Criteria: Acceptance criteria for TLAA related items are beyond the scope of current project.

Approach: TLAA (plant specific)

WCAP- 17096-NP December 2009 Revision 2

C-32 CE-ID: 9 Upper Internals Assembly Fuel Alignment Plate Category: Primary Applicability: All plants with core shrouds assembled with full-height shroud plates Degradation Effect: Cracking (fatigue)

Expansion Link: None Function: Provide fuel assembly alignment and support. Direct flow into upper internals.

Inspection Method: If fatigue life cannot be demonstrated by time-limited aging analysis (TLAA), enhanced visual (EVT-1) examination is required no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examination on a 10-year interval.

Coverage: Examination coverage to be defined by plant-specific fatigue analysis. See MRP-227 Figure 4-17.

Observable Effect: TLAA should be completed prior to inspection program. Normal rules for demonstrating fatigue life should be applied with updated projections of the number of load cycles.

Visual inspections for fatigue cracks along weld are required if sufficient fatigue life can not be demonstrated Failure Failure Mechanism: 1. Linkage of cracks from multiple origination sites leads to loss of integrity in fuel alignment plate.

2. Crack displacement causes misalignment of fuel assemblies.

Failure Effect 1. Loss of structural stability

2. Difficulty in loading fuel due to misalignment Failure Criteria: 1. Linkage of cracks that will create a critical flaw length

2. Linkage of cracks that will allow vertical displacement of a section of the fuel alignment plate Methodology Goal: Demonstrate that cracking of fuel alignment plate will not cause significant problems loading fuel.

Data Requirements: Stress analysis results for fatigue loading.

Analysis: Inspection of this item is required if sufficient fatigue life cannot be demonstrated by normal time-limited aging analysis (TLAA) procedures. A general outline of TLAA procedures is provided separately. The TLAA process evaluates potential fatigue crack initiation. As part of that evaluation, the stress amplitude and frequency must be estimated. If the TLAA indicates that crack initiation is possible, inspection of the indicated locations is required. Fatigue crack growth rates used in establishing acceptance criteria for the inspections should be based on the stress amplitudes and frequencies used in the TLAA.

Acceptance Criteria: Acceptance criteria for TLAA related items are beyond scope of current project.

Approach: TLAA (plant specific - applies to one utility)

WCAP- 17096-NP December 2009 Revision 2

C-33 CE-ID: 10 Control Element Assembly Instrument Guide Tubes Category: Primary Applicability: All plants with instrument guide tubes in the CEA shroud assembly Degradation Effect: Cracking (SCC, fatigue) that results in missing supports or separation at the welded joint Expansion Link: Remaining instrument guide tubes within the CEA shroud assemblies Function: Define path for insertion of in-core instrumentation.

Inspection Method: Visual (VT-3) examination, no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examination on a 10-year interval.

Plant-specific component integrity assessments may be required if degradation is detected and remedial action is needed.

Coverage: 100% of tubes in peripheral CEA shroud assemblies (i.e., those adjacent to the perimeter of the fuel alignment plate). See MRP-227 Figure 4-18.

Observable Effect: Missing or broken supports.

Failure Failure Mechanism: Cracking Failure Effect: 1. Potential loose parts

2. Inability to insert/withdraw instrumentation Failure Criteria: 1. Potential uncontained loose parts

2. Inability to maintain minimum in-core instrumentation Methodology Goal: Demonstrate ability to insert instrumentation.

Data Requirements: Instrumentation requirements for plant.

Analysis: 1. Evaluate stability of failed instrument guide tube. Any section that could potentially detach and become a loose part or otherwise interfere with plant operation should be removed or stabilized.

2. Any instrument guide tube with an observable crack will be assumed to have failed.

Acceptance Criteria: 1. Configuration of unfailed guide tubes should be sufficient to allow adequate core monitoring.

2. No margin is required for this item. If the instrumentation is functional at start-up, the plant can be operated.

Approach: Pass/Fail inspection with established minimum number of instrumentation tubes. Based directly on plant specifications.

WCAP- 17096-NP December 2009 Revision 2

C-34 CE-ID: 10.1 Control Element Assembly Remaining Instrument Guide Tubes Category: Expansion Applicability: All plants with instrument guide tubes in the CEA shroud assembly Degradation Effect: Cracking Expansion Link: Peripheral instrument guide tubes within the CEA shroud assemblies Function: Define path for insertion of in-core instrumentation.

Inspection Method: Visual (VT-3) examination, with initial and subsequent examinations dependent on the results of the instrument guide tubes examinations.

Coverage: 100% of tubes in CEA shroud assemblies.

See MRP-227 Figure 4-18.

Observable Effect: Missing or broken supports Failure Failure Mechanism: Cracking of attachment welds Failure Effect: 1. Potential loose parts

2. Inability to insert/withdraw instrumentation Failure Criteria: 1. Potential uncontained loose parts

2. Inability to maintain minimum in-core instrumentation Methodology Goal: Demonstrate ability to insert instrumentation.

Data Requirements: Instrumentation requirements for plant.

Analysis: 1. Evaluate stability of failed instrument guide tube. Any section that could potentially detach and become a loose part or otherwise interfere with plant operation should be removed or stabilized.

2. Any instrument guide tube with an observable crack will be assumed to have failed.

Acceptance Criteria: 1. Configuration of unfailed guide tubes should be sufficient to allow adequate core monitoring.

2. No margin is required for this item. If the instrumentation is functional at start-up, the plant can be operated.

Approach: Pass/Fail inspection with established minimum number of instrumentation tubes. Based directly on plant specifications. (See CE-ID: 10)

WCAP- 17096-NP December 2009 Revision 2

C-35 CE-ID: 11 Lower Support Structure Deep Beams Category: Primary Applicability: All plants with core shrouds assembled with full-height shroud plates Degradation Effect: Cracking (fatigue) Check for a detectable surface-breaking indication in the welds Expansion Link: None Function: Support core. Direct coolant flow into core.

Inspection Method: Enhanced visual (EVT- 1) examination, no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examination on a 10-year interval, if adequacy of remaining fatigue life cannot be demonstrated by TLAA.

Coverage: Examine beam-to-beam welds in the axial elevation from the beam top surface to four inches below. See MRP-227 Figure 4-19.

Observable Effect: Fatigue crack growth along welds at beams. Check for a detectable surface-breaking indication in the welds or beams.

Failure Failure Mechanism: Cracking Failure Effect: Loss of fuel assembly alignment Failure Criteria: No cracking that will cause displacement of fuel alignment pins Methodology Goal: Demonstrate stability of lower support structure.

Data Requirements: Potential fatigue loading and cycles.

Analysis: Inspection of this item is required if sufficient fatigue life cannot be demonstrated by normal time-limited aging analysis (TLAA) procedures. A general outline of TLAA procedures is provided separately. The TLAA process evaluates potential fatigue crack initiation. As part of that evaluation, the stress amplitude and frequency must be estimated. If the TLAA indicates that crack initiation is possible, inspection of the indicated locations is required. Fatigue crack growth rates used in establishing acceptance criteria for the inspections should be based on the stress amplitudes and frequencies used in the TLAA.

WCAP-1 7096-NP December 2009 Revision 2

C-36 CE-ID: 11 Lower Support Structure Deep Beams The general analysis of this structure would address the following issues:

I. The grid structure of the lower core support in these plants precludes catastrophic failure initiated by a single crack.

2. Cracking that will not result in failure of any beam or structural weld within the planned inspection interval should be acceptable.

• Assume crack initiation at most probable location as defined by TLAA.

• Evaluate crack depth (a)

- Determine crack growth rate consistent with stress amplitude and frequency used in TLAA.

- Project crack growth through planned inspection interval.

• Plot projected remaining ligament as a function of crack depth.

• Maximum acceptable crack size corresponds to projected remaining ligament = 0.

3. Additional margin against failure not required (catastrophic failure unlikely).

Acceptance Criteria: Acceptance criteria for TLAA related items are beyond scope of current project.

Approach: TLAA (plant specific - applies to one utility)

WCAP- 17096-NP December 2009 Revision 2

D-1 APPENDIX D FLOW CHARTS OF ILLUSTRATING EVALUATION METHODOLOGIES FOR COMBUSTION ENGINEERING-DESIGNED PLANTS CE Primary and Expansion Components CE-ID: 1 Core Shroud Assembly (Bolted) - Core Shroud Bolts CE-ID: 1.1 Core Shroud Assembly (Bolted) - Barrel-Shroud Bolts CE-ID: 1.2 Core Shroud Assembly (Bolted) - Core Support Column Bolts CE-ID: 2 Core Shroud Assembly (Welded) - Welds CE-ID: 2.1 Core Shroud Assembly (Welded) - Remaining Axial Welds CE-ID: 3 Core Shroud Assembly (Welded - Full Height) - Shroud Plates CE-ID: 3.1 Core Shroud Assembly (Welded - Full Height) - Axial Welds, Ribs and Rings CE-ID: 4 Core Shroud Assembly (Bolted) - Assembly CE-ID: 5 Core Shroud Assembly (Welded) - Assembly CE-ID: 6 Core Support Barrel Assembly - Upper (Core Support Barrel) Flange Weld CE-ID: 6.1 Core Support Barrel Assembly - Lower Core Barrel Flange CE-ID: 6.2 Core Support Barrel Assembly - Remaining Core Barrel Assembly Welds CE-ID: 6.3 Lower Support Structure - Core Support Column Welds CE-ID: 7 Core Support Barrel Assembly - Lower Flange Weld (Require TLAA- No Figure)

CE-ID: 8 Lower Support Structure - Core Support Plate (Require TLAA - No Figure)

CE-ID: 9 Upper Internals Assembly - Fuel Alignment Plate (Require TLAA - No Figure)

CE-ID: 10 Control Element Assembly - Instrument Guide Tubes CE-ID: 10.1 Control Element Assembly - Remaining Instrument Guide Tubes CE-ID: 11 Lower Support Structure - Deep Beams (Require TLAA - No Figure)

WCAP- 17096-NP December 2009 Revision 2

D-2 CE-ID: 1 Core Shroud Assembly -

Core Shroud Bolts WCAP- 17096-NP December 2009 Revision 2

D-3 CE-ID: 1.1 Core Shroud Assembly -

Barrel-Shroud Bolts WCAP- 17096-NP December 2009 Revision 2

D-4 CE-ID: 1.2 Core Shroud Assembly (Bolted)

Core support column bolts WCAP- 17096-NP December 2009 Revision 2

D-5 CE-ID: 2 Core Shroud Assembly (Welded)

Shroud Plate and CE-ID: 2.1 Core Shroud Assembly (Welded)

Remaining Axial Welds No Yes WCAP- 17096-NP December 2009 Revision 2

D-6 CE-ID: 3 Core Shroud Assembly (Welded-Full Height)

Shroud Plate and WCAP-17096-NP December 2009 Revision 2

D-7 CE-ID:4 Core Shroud Assemby (Bolted)

FMEA Layout

• -" *['~-- ----

,-'OSTO ........ -

ACTIVITY KEYINPUTI raS I DISPOSITION Failures Gray Record andg (Normal) IDocument Markings or gorShinyI Iu~s Develop Justification for Othe Continued Operation Record and Historical DocumentS lS IS

. Off-Normal

• e N L CniudOperation Develop Justification for S Continued Operation' g Plate IRecord and

• - -DocumentS Historical 1 SDocIPe CE-ID: 4 ý f~,pdo v-r-3 of Core Sroud New 1ý - Develop Justification for S r SContinued Operation (Bolted) ,

"**ckii~lg ircl kRecord and

• Failedca or Bol Document

• New Develop Justification for Continued Operation SNormal1 Record Plate No= and Conditions IDocument WCAP- 17096-NP December 2009 Revision 2

D-8 CE-tD:5 Core Shroud Assemby (Welded)

FMEA Layout ACTIVITY KEY INPUTS DISPOSITION Failures Y

,Normal Gray= *D~uet Record and Markings orocument Shiny "

Iug r Develop Justification for Continued Operation Other Ma Record and SeparatiDocument N Develop Justification for Continued Operationf e- Plate RCondtons1isicalecord and

-P Document S CE-ID: 5 Warped

=_* or on-fo VT-3 of Crackedw Develop Justification for S Core Shroud -Continued Operation (Welde) " *)* J"Re~cord and Issues - -mn~

SOther I IS Develop Justificationýfor Ne New l *" Continued Operation SNormal1 Rerd Plate an ConditionsDocument WCAP- 17096-N P December 2009 Revision 2

D-9 CE-ID:6 Core Support Barrel Assembly - Upper Flange Weld Option 1 OD Inspection Yes No Determine applied stress intensity factor, Kiapp, for projected length under faulted condition Note: Flaw dimension Lc is limit where growth prohibited by stress distribution.

WCAP- 17096-NP December 2009 Revision 2

D-10 CE-ID: 6 Core Support Barrel Assembly - Upper Flange Weld Option 2: ID Inspection WCAP- 17096-NP December 2009 Revision 2

D-I I CE-ID: 6.1 Core Support Barrel Assembly Lower Flange and WCAP- 17096-NP December 2009 Revision 2

D- 12 CE-ID: 6.3 Lower Support Structure Core Support Column Welds WCAP- 17096-NP December 2009 Revision 2

D-13 CE-ID:10 &10.1 Control Element Assembly Instrument Guide Tubes WCAP- 17096-NP December 2009 Revision 2

E-1 APPENDIX E ACCEPTANCE CRITERIA METHODOLOGY AND DATA REQUIREMENTS FOR WESTINGHOUSE COMPONENTS INCLUDED IN MRP-227 Westinghouse Primary and Expansion Components W-ID: 1 Control Rod Guide Tube Assembly - Guide Pates (Cards)

W-ID: 2 Control Rod Guide Tube Assembly - Lower Flange Welds W-ID: 2.1 Lower Support Assembly - Lower Support Column Bodies (Cast)

W-ID: 2.2 Bottom-mounted Instrumentation (BMI) System - BMI Column Bodies W-ID: 3 Core Barrel Assembly - Upper Core Barrel Flange Weld W-ID: 3.1 Core Barrel Assembly - Core Barrel Flange, Core Barrel Outlet Nozzles, Lower Core Barrel Flange Weld W-ID: 3.2 Lower Support Assembly - Lower Support Columns (non cast)

W-ID: 4 Baffle-former Assembly - Baffle-edge Bolts W-ID: 5 Baffle-former Assembly - Baffle-Former Bolts W-ID: 5.1 Core Barrel Assembly - Barrel-Former Bolts W-ID: 5.2 Lower Support Assembly - Lower Support Column Bolts W-ID: 6 Baffle-Former Assembly - Assembly W-ID: 7 Alignment and Interfacing Components - Internal Hold-down Spring W-ID: 8 Thermal Sleeve Assembly - Thermal Shield Flexures WCAP- 17096-NP December 2009 Revision 2

E-2 W-ID: 1 Control Rod Guide Tube Assembly Guide Plates (Cards)

Category: Primary Applicability: All plants Degradation. Effect: Loss of Material (Wear)

Expansion Link: None.

Function: The control rod guide tube assembly provides alignment and insertion path for the control rods through the upper internals. Guide cards provide alignment and insertion path for control rod assemblies and support the control rods when withdrawn.

Inspection Method: Visual (VT-3) examination no later than 2 refueling outages from the beginning of the license renewal period, and no earlier than two refueling outages prior to the start of the license renewal period. Subsequent examinations are required on a 10-year interval.

Coverage: 20% examination of the number of CRGT assemblies, with all guide cards within each selected CRGT assembly examined.

See MRP-227 Figure 4-20 Observable Effect: Observation of wear requires internal visual inspections of guide tube assemblies.

Westinghouse has established procedure for quantifying wear based on calibrated visual exams for the PWROG. The Westinghouse procedures meet and exceed the VT-3 requirements.

The VT-3 inspections should be able to identify ligaments on inner guidance holes.

Failure Failure Mechanism: The guidance holes in the guide cards are distorted by wear (loss of material). Largest amounts of wear typically observed in lowest guide card levels.

Failure Effect: Guidance hole wear can cause lack of alignment. Lack of alignment may cause a degradation of control rod drop times. In the worst case scenario, rod may jam and prevent insertion.

Failure Criteria: Leading indicator of failure is considered to be observation of sharp tip at inner guide card slots.

• Wear such that rod may escape is currently considered as failure of guide card.

• Failure requires control rod to wear through ligament in guide card.

Methodology Goal: Rod must be restrained to guidance hole in card.

Data Requirements: Guide card wear model Material properties Current geometry Wear trend Vertical and horizontal location Maintenance practices WCAP-1 7096-NP December 2009 Revision 2

E-3 W-ID: 1 Control Rod Guide Tube Assembly Guide Plates (Cards)

Control Rod Insertion Data (Historical)

Analysis: Two stages of wear:

A. Wear through full ligament. Can observe enlargement of guide card hole, but wear does not extend to inside surface of guide card.

B. Wear area intersects inner surface of guide card, but wear slot still too narrow to allow escape of control rod.

Unworn Stage A Stage B During Stage A, one should be able to observe a flat, unwom surface on slots in inner hole. Upon transition from Stage A to Stage B, there is no observable slot. Sharp point observed where wear area intersects inner surface of guide card.

Bounding calculation for wear life.

1. Must assume "typical" wear patterns as previously observed in PWROG program.

2. Calculate wear volume (Va) at transition from Stage A to Stage B.

3. Calculate wear volume (Vb) at point where width of wear area at guide tube inner surface is equivalent to control rod diameter.

4. Calculate fraction wear f = Va/Vb.

5. Calculate remaining wear life T = (1/f - 1)Tcur (Tcur = current operating time)

Inspection interval must be less than remaining wear life.

Acceptance Criteria: Require control rod to be captured in guide card hole.

The unworn section of guide card slot must be observable at all inner guide tube holes at each guide card level.

Demonstrate wear remains in Stage A (see analysis).

Approach: Generic work ongoing under PWROG program Validate and/or modify linear volumetric wear rate model Potential extension Alternative justification that allows wear through ligament in one or more cards WCAP- 17096-NP December 2009 Revision 2

E-4 W-ID: 2 Control Rod Guide Tube Assembly Lower Flange Welds Category: Primary Applicability: All plants Degradation Effect: Cracking (SCC, fatigue)

Expansion Link: Bottom-mounted instrumentation (BMI) column bodies, Lower support column bodies (cast)

Function: The control rod guide tube assembly provides alignment and insertion path for control rods through upper internals. The lower flange welds retain the structural alignment of the component. Guide tubes must maintain rod stability in normal and LOCA transients.

Inspection Method: Enhanced visual (EVT-1) examination to determine the presence of crack-like surface flaws in flange welds no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

Coverage: 100% of outer (accessible) CRGT lower flange weld surfaces and adjacent base metal.

See MRP-227 Figure 4-21.

Observable Effect: Any individual weld with observed crack must be assumed to have failed.

The vertical beam portion appears to be out of position.

Failure Failure Mechanism: Flow in the upper head applies bending moment to control rod guide tube assembly.

Maximum bending stresses tend to occur near top of continuous guidance section.

Stresses may lead to formation of SCC or fatigue cracks. Weld cracking may lead to loss of stiffness in guide tube assembly and loss of support capability.

Failure Effect: Loss of structural stability. Excessive deflection could impede control assembly insertion.

Failure Criteria: Design limits on the CRGT assembly are generally expressed as a maximum allowable load, which is determined based on the assembly compliance. This analysis implies a maximum allowable deflection. Interference between the guide cards and the guide tubes occur when the deflection exceeds this limit.

Methodology Goal: Stiffness of assembly with failed welds must be sufficient to maintain allowable deflections when LOCA and SSE loads are applied. Allowable load on control rod guide tube assembly is defined by empirical testing.

Data Requirements: Loads Finite element model of lower CRGT assembly to evaluate weld failures calibrated to benchmark data Analysis: 1. Determine design basis assumptions for CRGT assembly (maximum allowable load, assembly compliance).

2. Lower section must be modeled in detail, upper sections may be treated as large beams.

3. Calibrate FEA model and boundary constraints against design basis assumptions.

WCAP- 17096-NP December 2009 Revision 2

E-5 W-ID: 2 Control Rod Guide Tube Assembly Lower Flange Welds

4. Remove test pattern of welds.

5. Run FEA.

6. If deflection is greater than limit in Step 1, pattern is not acceptable.

7. Iterate Steps 4-6 to create library of acceptable and unacceptable patterns.

8. Match patterns to field observations assuming that any weld with flaw has failed.

9. Should be able to observe sufficient number of welds to demonstrate that assembly is acceptable.

Acceptance Criteria: This acceptance criteria is based on a minimum number of welds that must continue to function (without cracking) to allow scramming of the control rods in the event of combined LOCA and SSE.

II n Top Welt fin II fin SFailed WeliA Bottom Welds U

r Intact Weld Approach: Plant-specific analysis due to large variety of sizes and designs. There may be some potential for smaller plant groupings.

WCAP-17096-NP December 2009 Revision 2

E-6 W-ID: 2.1 Lower Support Assembly Lower Support Column Bodies (Cast)

Category: Expansion Applicability: All plants Degradation Effect: Cracking (IASCC) including the detection of fractured support columns Expansion Link: Control rod guide tube (CRGT) lower flanges Function: The lower support columns provide the structural link between the lower core plate and the lower support structure. The supports are required to keep the lower core plate from deforming during operation.

Inspection Method: Visual (EVT- 1) examination Coverage: 100% of accessible support columns See MRP-227 Figure 4-34.

Observable Effect: Fracture Potential for core tilt Control rod insertion problems Failure Failure Mechanism: The upper sections of the core supports may experience neutron fluences above the threshold for IASCC. The cast components are considered separately because there is a concern that they may be more sensitive to irradiation. Although stresses in columns are primarily compressive, bending stresses or the design of the attachment may produce localized regions of tensile stress.

Failure Effect: Displacement of lower core plate Failure Criteria: Must maintain sufficient number of intact support columns to assure dimensional stability of lower core plate.

Methodology Goal: Establish minimum acceptable pattern of core support columns. Evaluation of cast components should consider potential effect of thermal embrittlement in addition to irradiation embrittlement.

Data Requirements: 0 Loads on lower core plate 0 Constitutive model for stainless steel properties as a function of irradiation and thermal aging

• Displacement tolerances on lower core plate Analysis: 1. Establish minimum functional requirements and number of core support columns to maintain structure and functional stability.

A. During normal operation system of support columns should resist core plate deformation due to mechanical or thermal loading. Core plate requirements for "flatness" and fuel assembly alignment.

B. During limiting accident transient system must maintain structural integrity.

WCAP- 17096-NP December 2009 Revision 2

E-7 W-ID: 2.1 Lower Support Assembly Lower Support Column Bodies (Cast)

2. Support column analysis assumptions.

A. Assume any column with crack in main body to have failed.

B. Assume any column with a crack in attachment device or bolt to result in failure of the attachment.

3. Structural model of lower support structure.

• FEA model of lower support structure that includes support columns and lower core plate. Model should be capable of removing individual column or breaking attachment to lower core plate. Would require multiple iterations to establish "minimum acceptable patterns" of core support columns and support column bolts.

4. Structural model must be run for functional requirements A and B.

5. Determine margin for additional failures.

A. Assume number of failures in next 10 years is equal to number observed to date.

N = # of support columns Nf = # of observed flawed columns Nreq = # of columns in relevant minimum pattern Margin = N-Nreq Acceptance Criteria: Require that no more of 1/2 of columns in margin are failed:

Nf < (N-Nreq)/2 Approach: Generic program to share first-of-a-kind effort. (See W-ID: 2-1)

• Pilot analysis of lower support structure to identify critical issues.

Expect final acceptance based on plant-specific analysis.

WCAP- 17096-NP December 2009 Revision 2

E-8 W-ID: 2.2 Bottom-mounted Instrumentation System Bottom-mounted Instrumentation (BMI) Column Bodies Category: Expansion Applicability: All plants Degradation Effect: Cracking (fatigue) including the detection of completely fractured column bodies Expansion Link: Control rod guide tube (CRGT) lower flanges Function:

• The BMI columns define the path for flux thimbles to be inserted into the fuel assemblies.

Flux thimbles are normally withdrawn prior to refueling and re-inserted at end of refueling.

The plant must maintain a required number of functioning flux thimbles for core mapping.

Inspection Method: Visual (VT-3) examination of BMI column bodies as indicated by difficulty of insertion/withdrawal of flux thimbles. Flux thimble insertion/withdrawal to be monitored at each inspection interval.

Coverage: 100% of BMI column bodies for which difficulty is detected during flux thimble insertion/withdrawal.

See MRP-227 Figure 4-35.

Observable Effect:

• Fracture should be readily visible 0 Large loose parts Skewed flow

• Weakened support Failure Failure Mechanism: The BMI columns may be subject to fatigue due to either thermal fatigue or flow induced vibrations.

Failure Effect: Inability to insert flux thimbles. This effect would be noted during refueling outage.

Consequences of failure during ensuing operating period are believed to be minimal.

Failure Criteria:

• The plant must maintain a required number of functioning flux thimbles for core mapping.

• Any BMI column with an observable crack will be assumed to have failed.

• The primary pressure boundary must be intact.

Methodology Goal: Configuration of unfailed BMI columns should be sufficient to allow required flux mapping. (Installation of WINCISETM may obviate the need for the entire BMI system.)

Data Requirements: Criteria should be part of plant technical specifications.

Analysis: Evaluate stability of failed BMI Column. Any section that could potentially detach and become a loose part or otherwise interfere with plant operation should be removed or stabilized.

WCAP- 17096-NP December 2009 Revision 2

E-9 W-ID: 2.2 Bottom-mounted Instrumentation System Bottom-mounted Instrumentation (BMI) Column Bodies Acceptance Criteria: Plant must have minimum number of unfailed BMI assemblies to allow flux-mapping at startup.

Approach: Pass/Fail inspection with established minimum number of instrumentation tubes. Based directly on plant specifications.

WCAP- 17096-NP December 2009 Revision 2

E-10 W-ID: 3 Core Barrel Assembly Upper Core Barrel Flange Weld Category: Primary Applicability: All plants Degradation Effect: Cracking (SCC)

Expansion Link: Remaining core barrel welds (core barrel flange, core barrel outlet nozzles, lower core barrel flange weld), lower support column bodies (non cast)

Function: Primary core support structure.

Inspection Method: Periodic enhanced visual (EVT- 1) examination, no later than 2 refueling outages from the beginning of the license renewal period and subsequent examination on a 10-year interval.

Coverage: 100% of one side of the accessible surfaces of the selected weld and adjacent base metal.

See MRP-227 Figure 4-22.

Observable Effect: Stress corrosion crack along seam weld.

Failure Failure Mechanism: SCC Failure Effect: Potential loss of core support.

Failure Criteria: Actively growing through-wall flaws require mitigation. Require demonstration that flaw growth is arrested or limited to surface.

An existing through-wall flaw may be acceptable if condition and shape indicate that it is a non-growing fabrication flaw.

Methodology Goal: Due to the high fracture toughness of unirradiated stainless steel, the core barrel is a highly flaw tolerant structure and flaw sizes are expected to be very large. However, the core barrel is a critical support structure. Flaw growth in this component is outside the range of normal expectations. Therefore, it has been assumed that the presence of any actively growing through-wall crack would require repair or other mitigation. The goal of the calculation is to demonstrate the crack is stable or not likely to grow through wall.

Data Requirements: 1. Operating loads

2. K solutions for range of expected crack shapes (lengths and depths)

3. SCC crack growth rate curves

4. Fatigue crack growth rate curve (as backup)

Analysis: Option 1. Observation on OD of core barrel Step 1. Determine stress distribution through core barrel thickness for normal operating conditions (expect peak stress at vessel OD).

Step 2. Obtain stress intensity factor solution for part-through-wall crack as function of surface length (L) and depth (a).

Step 3. Short cracks will be constrained by the stress distribution in the WCAP- 17096-NP December 2009 Revision 2

E-11 W-ID: 3 Core Barrel Assembly Upper Core Barrel Flange Weld barrel wall. Define the maximum constrained crack length as Lc.

Step 4. OD crack observation is acceptable if L < Lc.

Step 5 If L > Lc, then must perform UT to determine crack depth (a).

Step 6. Crack is acceptable if K corresponding to a and Lc is less than 20 ksi-in^1/2.

Step 7. All remaining cracks require specific flaw-tolerance analysis.

Option 2. Observation of flaw on ID of core support barrel Step 1. If flaw on ID is smaller than the length (Lc) defined in Option 1, visually examine the OD surface corresponding to the ID flaw to determine if it is OD-initiated. Crack is acceptable if not through-wall.

Step 2. For a through-wall flaw, apply the OD flaw acceptance criteria from Option 1.

Step 3. All remaining cracks require a geometry-specific flaw-tolerance analysis.

Option 3. Observation of crack on ID of core support barrel Step 1. If flaw on ID is smaller than the length (Lc) defined in Option 1, perform UT exam to determine if the crack is through-wall. Crack is acceptable if not through-wall.

Step 2. For a through-wall flaw, apply the OD flaw acceptance criteria from Option 1.

Step 3. All remaining cracks require a geometry-specific flaw-tolerance analysis.

Acceptance Criteria: Demonstrate that crack is not actively growing or limited to surface as indicted by analysis.

Approach: Plant-specific analysis.

Ginna provides pilot plant experience for the creation of generic acceptance criteria.

May be able to group plants by design.

WCAP- 17096-NP December 2009 Revision 2

E- 12 W-ID: 3.1 Core Barrel Assembly Core Barrel Flange, Core Barrel Outlet Nozzles, Lower Core Barrel Flange Weld Category: Expansion Applicability: All plants Degradation Effect: Cracking (SCC, fatigue)

Expansion Link: Upper core barrel flange weld Function: Primary core support structure Inspection Method: Enhanced visual (EVT-1) examination, with initial examination and re-examination frequency dependent on the examination results for upper core barrel flange Coverage: 100% of one side of the accessible surfaces of the selected weld and adjacent base metal See MRP-227 Figure 4-22.

Observable Effect: Cracking along line of weld.

Failure Failure Mechanism: SCC, fatigue Failure Effect: Potential loss of core support Failure Criteria: Actively growing through-wall flaws require mitigation. Require determination of crack growth mechanism.

An existing through-wall flaw may be acceptable if condition and shape indicate that it is a non-growing fabrication flaw.

Methodology Goal: Demonstrate that cracking mechanism is understood and projected crack growth is limited.

Data Requirements: 1. Operating loads

2. K solutions for range of expected crack shapes (lengths and depths)

3. SCC crack growth rate curves

4. Fatigue crack growth rate curve (as backup)

Analysis: 1. Flaws in core barrel above the baffle section will be evaluated assuming active crack growth mechanisms are SCC and fatigue.

2. Flaws in the beltline region of the core barrel (care baffle section) will be evaluated assuming active growth mechanisms are IASCC and fatigue.

3. A fluence estimate at the flaw location is required for all flaws in the beltline region.

4. Normal operating and fatigue loads will be established for core barrel at this location.

5. Determine stress intensity factors for a through-wall crack.

6. Use appropriate crack growth rate models (SCC or IASCC and fatigue) to estimate crack growth rate.

WCAP- 17096-NP December 2009 Revision 2

E- 13 W-ID: 3.1 Core Barrel Assembly Core Barrel Flange, Core Barrel Outlet Nozzles, Lower Core Barrel Flange Weld

7. If crack growth rate is consistent with observed flaw size:

• Project flaw size through inspection interval using crack growth rate estimate.

• Determine loads during limiting transient.

• Determine stress intensity factor for through-wall crack of projected length

• For low fluence region assume Klc = 150 ksi-in^l/2

" For beltline region determine lower bound toughness based on fluence estimate.

• If stress intensity factor during transient is less than fracture toughness, flaw is acceptable.

" If stress intensity factor during transient is greater than fracture toughness, proceed to Step 8

8. If crack growth rate is too low to explain existence of observed crack or flaw not acceptable by Step 7:

Determine crack depth

• If crack depth small compared to barrel thickness (< xx inches) then crack is acceptable.

" If crack depth large compared to barrel thickness, the crack is rapidly growing and a detailed analysis is required.

Acceptance Criteria: Current crack size is explainable by known crack growth rate laws and limited crack growth is projected.

Approach: Plant-specific analysis.

Require flaw tolerance handbook/methodology based on flaw location and direction.

MRP-2 10 may have limited relevance.

WCAP- 17096-NP December 2009 Revision 2

E- 14 W-ID: 3.2 Lower Support Assembly Lower Support Column Bodies (Non Cast)

Category: Expansion Applicability: All plants Degradation Effect: Cracking (JASCC)

Expansion Link: Upper core barrel flange weld Function: The lower support columns provide the structural link between the lower core plate that supports the fuel assemblies and the relatively thick lower support forging (or in a limited number of cases casting.) The supports are required to keep the lower core plate from deforming during operation.

Inspection Method: Enhanced visual (EVT- 1) examination, with initial examination and re-examination frequency dependent on the examination results for upper core barrel flange weld.

Coverage: 100% of accessible surfaces See MRP-227 Figure 4-34.

Observable Effect: e Fracture

• Potential for core tilt

• Control rod insertion problems Failure Failure Mechanism: The upper sections of the core supports may experience neutron fluences above the threshold for IASCC. Although the main stresses in the support is expected to be compressive, bending stresses or the design of the attachment may produce localized regions of tensile stress.

Failure Effect: Displacement of lower core plate Failure Criteria: Must maintain sufficient number of intact support columns to assure dimensional stability of lower core plate.

Methodology Goal: Establish minimum acceptable pattern of core support columns.

Data Requirements: Loads on lower core plate Constitutive model for stainless steel properties as a function of irradiation and thermal aging.

Displacement tolerances on lower core plate Analysis: 1. Establish minimum functional requirements and number of core support columns to maintain structure and functional stability.

A. During normal operation system of support columns should resist core plate deformation due to mechanical or thermal loading. Core plate requirements for "flatness" and fuel assembly alignment.

B. During limiting accident transient system must maintain structural integrity.

WCAP- 17096-NP December 2009 Revision 2

E- 15 W-ID: 3.2 Lower Support Assembly Lower Support Column Bodies (Non Cast)

2. Support column analysis assumptions.

A. Assume any column with crack in main body to have failed.

• B. Assume any column with a crack in attachment device or bolt to result in failure of the attachment.

3. Structural model of lower support structure.

. FEA model of lower support structure that includes support columns and lower core plate. Model should be capable of removing individual column or breaking attachment to lower core plate. Would require multiple iterations to establish "minimum acceptable patterns" of core support columns and support column bolts.

4. Structural model must be run for functional requirements A and B.

5. Determine margin for additional failures.

A. Assume number of failures in next 10 years is equal to number observed to date.

N = # of Support Columns Nf = # of Observed Flawed Columns Nreq = # of columns in relevant minimum pattern Margin = N - Nreq Acceptance Criteria: Require that no more of 1/2 of columns in margin are failed:

Nf < (N - Nreq)/2 Approach: Generic program to share first-of-a-kind effort. (See W-ID: 2-1)

• Pilot analysis of lower support structure to identify critical issues.

• Expect final acceptance based on plant-specific analysis.

WCAP-17096-NP December 2009 Revision 2

E- 16 W-ID: 4 Baffle-former Assembly Baffle-edge Bolts Category: Primary Applicability: All plants with baffle-edge bolts Degradation Effect: Cracking (IASCC, fatigue) that results in Expansion Link: None Function: The baffle-edge bolts provide the baffle-plate to baffle plate attachment along the seam between plates. The edge bolts prevent gaps between plants that can result in baffle-jetting damage to peripheral fuel assemblies.

Studies have demonstrated that baffle edge bolts are not required to maintain the structural integrity of the baffle.

Inspection Method: Visual (VT-3) examination, with baseline examination between 20 and 40 EFPY and subsequent examinations on a 10-year interval.

Coverage: Bolts and locking devices on high fluence seams. 100% of components accessible from core side.

See MRP-227 Figure 4-23.

Observable Effect: Failure of bolt or locking device as listed under inspection.

Failure Failure Mechanism: Analysis has shown that differential thermal expansion and swelling can cause plastic deformation of edge bolts. These bolts are in high radiation locations and there is a significant potential failure due to IASCC.

Failure modes considered should include:

• Broken or missing locking devices

• Protruding bolt heads

• Missing bolts or bolt heads Failure Effect: In plants with downward coolant flow in the region between the baffle and the former, failure may contribute to baffle jetting.

Primary concerns are loose parts generation and interference with fuel.

Failure Criteria: All bolts and locking devices should be in place and undamaged. FMEA should be completed prior to analysis to identify potential observations. Pre-planned responses to be implemented.

Methodology Goal: A plant-specific plan should be developed for evaluating and mitigating the potential relevant conditions. The evaluation should consider any previously reported observations.

Data Requirements: FMEA results Analysis: Prepare for examination by conducting a failure modes and effects analysis (FMEA) to identify full range of potential relevant observations prior to inspection.

WCAP- 17096-NP December 2009 Revision 2

E-17 W-ID: 4 Baffle-former Assembly Baffle-edge Bolts Acceptance Criteria: Determined by FMEA Approach: FMEA should address plant-specific practices and priorities. Some generic work possible to outline issues and options to be addressed in FMEA.

WCAP- 17096-NP December 2009 Revision 2

E-18 W-ID: 5 Baffle-former Assembly Baffle-former Bolts Category: Primary Applicability: All plants Degradation Effect: Cracking (IASCC, fatigue)

Expansion Link: Lower support column bolts, barrel-former bolts Function: The baffle-former bolts attach the baffle plates to the formers.

Inspection Method: Baseline volumetric (UT) examination between 25 and 35 EFPY, with subsequent examination after 10 to 15 additional EFPY to confirm stability of bolting pattern.

Re-examination for high-leakage core designs requires continuing examinations on a 10-year interval.

Coverage: 100% of accessible bolts or as supported by plant-specific justification. Heads accessible from the core side. UTaccessibility may be affected by complexity of head and locking device designs.

See MRP-227 Figures 4-23 and 4-24.

Observable Effect: UT will detect bolts with large cracks (approx. 30%) through of cross-sectional area.

Fractured bolts should be captured by locking devices - no visible indication.

Failure Failure Mechanism: Known IASCC cracking of similar highly irradiated bolts has been reported.

Failure Effect: Loss of structural stability Failure Criteria: Require a minimum bolting pattern Methodology Goal: Must demonstrate that projected number of additional bolt failures will not threaten minimum pattern prior to next scheduled inspection.

Data Requirements: a Loads Bolting patterns

• Baffle design

• Fast neutron (dpa) distribution in core shroud

• Projected bolt failure rate

• Minimum bolting pattern analysis Analysis: The observed pattern of failed bolts must meet the pre-defined acceptable bolt pattern and have a reasonable margin to protect against additional failures during the inspection interval. The margin is defined in terms of the number of intact bolts beyond the number required for the minimum bolting pattern. The margin (M) at any time is simply:

M = N - Nreq - Nf where N = total number of baffle-former bolts Nreq = number of baffle-former bolts in minimum acceptable pattern Nf = number of failed bolts WCAP- 17096-NP December 2009 Revision 2

E-19 W-ID: 5 Baffle-former Assembly Baffle-former Bolts Assuming that there are no failed bolts at the beginning of life, the initial margin is simply: (N - Nreq). For operation through the next 10-15 EFPY interval, require that no more than 50% of initial margin be consumed at the time of the first inspection.

Acceptance Criteria: 1. Observed pattern of unfailed bolts meets pre-defmed acceptance criteria

2. Less than 50% of initial margin consumed Nf < (N - Nreq)/2 Approach: Generic work completed in previous PWROG program WCAP- 17096-NP December 2009 Revision 2

E-20 W-ID: 5.1 Core Barrel Assembly Barrel-former Bolts Category: Expansion Applicability: All plants Degradation Effect: Cracking (IASCC, fatigue)

Expansion Link: Baffle-former bolts Function: Maintain structural integrity of baffle-former-barrel structure.

Inspection Method: Volumetric (UT) examination, with initial and subsequent examinations dependent on results of baffle-former bolt examinations.

Coverage: 100% of accessible bolts. Accessibility may be limited by presence of thermal shields or neutron pads.

See MRP-227 Figure 4-23.

Observable Effect: UT will detect bolts with large cracks (approx. 30%) through the cross sectional area Failure Failure Mechanism: Cracking Loss of bolt pre-load due to irradiation induced stress relaxation may exacerbate fatigue issue in aging plants Failure Effect: Potential for flow induced vibration due to loss of bolting constraint.

Loss of structural stability Failure Criteria: UT indications Methodology Goal: Must demonstrate a minimum bolting pattern.

Data Requirements:

• Loads/displacements

• Bolting pattems

• Baffle-former = barrel design

• Fast neutron (dpa) distribution in core barrel

• Projected bolt failure rate

• Minimum bolting pattem analysis Analysis: Procedures for establishing acceptable bolting pattems for the barrel-to-former bolts have been established in [13]. This methodology has been reviewed and accepted by the NRC in a Safety Evaluation issued in 1998 (TAC No. MAI 152). The PWROG has developed minimum acceptable bolting patterns for all Westinghouse designed plants in the United States. In some cases, a plant-specific bolting pattern evaluation may produce a less restrictive result.

WCAP- 17096-NP December 2009 Revision 2

E-21 W-ID: 5.1 Core Barrel Assembly Barrel-former Bolts The observed pattern of failed bolts must meet the pre-defined acceptable bolt pattern and have a reasonable margin to protect against additional failures during the inspection interval. The margin is defined in terms of the number of intact bolts beyond the number required for the minimum bolting pattern. The margin (M) at any time is simply:

M = N - Nreq - Nf where N = total number of barrel-former bolts Nreq = number of barrel-former bolts in minimum acceptable pattern Nf = number of failed bolts Assuming that there are no failed bolts at the beginning of life, the initial margin is simply: (N - Nreq). For operation through the next 10-15 EFPY interval, require that no more than 50% of initial margin be consumed at the time of the first inspection.

Acceptance Criteria: 1. Observed pattern of unfailed bolts meets pre-defmed acceptance criteria

2. Less than 50% of initial margin consumed Nf < (N - Nreq)/2 Approach: Generic work completed in previous PWROG program December 2009 17096-NP WCAP- 17096-NP December 2009 Revision 2

E-22 W-ID: 5.2 Lower Support Assembly Lower Support Column Bolts Category: Expansion Applicability: All plants Degradation Effect: Cracking (IASCC, fatigue)

Expansion Link: Baffle-former bolts Function: The lower support column bolts attach the support columns to the lower core plate.

Although the bolts do not directly support the weight of the core, they help maintain the flatness and integrity of the lower support plate.

Inspection Method: Volumetric (UT) examination, with initial and subsequent examinations dependent on results of baffle-former bolt examinations.

Coverage: 100% of accessible bolts or as supported by plant-specific justification.

See MRP-227 Figures 4-32 and 4-33.

Observable Effect: Failed UT inspection Failure Failure Mechanism: Cracking Failure Effect: Displacement of lower core plate Failure Criteria: Assume failure of bolt results in loss of attachment between support column and lower core plate.

Methodology Goal: Establish functional requirements for core support columns.

A. During normal operation system of support columns should resist core plate deformation due to mechanical or thermal loading. Core plate requirements for "flatness" and fuel assembly alignment.

B. During limiting accident transient system must maintain structural integrity.

Data Requirements:

• Loads on lower core plate Displacement tolerances on lower core plate Analysis: 1. Establish functional requirements for core support columns.

A. During normal operation system of support columns should resist core plate deformation due to mechanical or thermal loading. Core plate requirements for "flatness" and fuel assembly alignment.

B. During limiting accident transient system must maintain structural integrity.

2. Structural model of lower support structure.

FEA model of lower support structure that includes support columns and lower core plate. Model should be capable of removing individual column or breaking attachment to lower core plate. Would require multiple iterations to establish "minimum acceptable patterns" of core support columns and support column bolts.

3. Structural model must be run for functional requirements A and B.

WCAP- 17096-NP December 2009 Revision 2

E-23 W-ID: 5.2 Lower Support Assembly Lower Support Column Bolts

4. Determine margin for additional failures.

A. Assume number of failures in next 10 years is equal to number observed to date.

N = # of Support Columns Nf = # of Observed Flawed Columns Nreq = # of columns in relevant minimum pattern Margin = N - Nreq Acceptance Criteria: Nf < (N - Nreq)/2 Approach: Generic program to share first-of-a-kind effort. (See W-ID: 2-1)

Pilot analysis of lower support structure to identify critical issues.

Expect final acceptance based on plant-specific analysis.

WCAP- 17096-NP December 2009 Revision 2

E-24 W-ID: 6 Baffle-former Assembly Assembly Category: Primary Applicability: All plants Degradation Effect: Distortion (void swelling), or cracking (IASCC) that results in Expansion Link: None Function:

• Provide support, guidance, and protection for the reactor core

• Provide ana~aewav for the distrihbution of the reactsnr eoolan~t flow to the reactor core Provide gamma and neutron shielding for the reactor vessel Inspection Method: Visual (VT-3) examination to check for evidence of distortion, with baseline examination between 20 and 40 EFPY and subsequent examinations on a 10-year interval.

Coverage: Core side surface as indicated See MRP-227 Figures 4-24, 4-25, 4-26, and 4-27.

Observable Effect: Degradation of general condition as described above Failure Failure Mechanism: Void swelling, IASCC Failure Effect: 1. Interference with fuel assemblies

2. Obstruction of coolant flow

3. Loose parts generation

4. Distortion/misalignment of core

5. Local temperature peaks

6. Degradation of control rod insertability

7. Baffle jetting Failure Criteria: No relevant observations Methodology Goal: A plant-specific plan should be developed for evaluating and mitigating the potential relevant conditions. The evaluation should consider any previously reported observations.

Data Requirements: 1. Baseline data on previous visual examinations of baffle-former assembly

2. Loose parts monitoring data Analysis: Prepare for examination by conducting a failure modes and effects analysis (FMEA) to identify full range of potential relevant observations prior to inspection. Failure mechanisms considered should include:

• Broken or missing locking devices

• Protruding bolt heads Missing bolts or bolt heads WCAP- 17096-NP December 2009 Revision 2

E-25 W-ID: 6 Baffle-former Assembly Assembly

• Distortion or displacement of baffle plates

• Gross cracking of baffle plates

• Gaps at plate joints

• Interaction with fuel assemblies

• Historical record Acceptance Criteria: Determined by FMEA Approach: FMEA should address plant-specific practices and priorities. Some generic work possible to outline issues and options to be addressed in FMEA.

WCAP-17096-NP December 2009 Revision 2

E-26 W-ID: 7 Alignment and Interfacing Components Internals Hold-down Spring Category: Primary Applicability: All plants with 304 stainless steel hold-down springs Degradation Effect: Stress Relaxation Expansion Link: None Function: Provide hold-down forces for core internals.

Retain internals in proper alignment to the core.

Inspection Method: Direct measurement of spring height within three cycles of the beginning of the license renewal period. If the first set of measurements is not sufficient to determine life, spring height measurements must be taken during the next two outages to extrapolate the expected spring height to 60 years.

Coverage: Measurements should be taken at several points around the circumference of the spring, with a statistically adequate number of measurements at each point to minimize uncertainty.

See MRP-227 Figure 4-28.

Observable Effect: Reduced height of core hold-down spring. Repeated measurements should indicate progressive reduction in height from cycle-to-cycle. Wear surfaces may also exhibit evidence of galling.

Failure Failure Mechanism: Stress relaxation Failure Effect:

• Loss of hold-down forces may lead to vibration and wear in lower internals 0 Long term stress relaxation Failure Criteria: Failure to maintain hold-down force through next inspection cycle.

Methodology Goal: Remaining spring force must meet requirements for core hold-down forces.

Data Requirements:

• Historical information on spring height (project rate of relaxation)

• Effective spring constant

• Necessary hold-down force (plant specific)

• Current spring height

• Degradation (trending)

Analysis: Need to construct creep-stress relaxation model to define bounding (high relaxation) behavior.

• Material properties (stiffness, creep)

• History (transients, creep)

• Geometry

• Force profile WCAP- 17096-NP December 2009 Revision 2

E-27 W-ID: 7 Alignment and Interfacing Components Internals Hold-down Spring Acceptance Criteria: Relaxation of hold-down spring must be above bounding prediction.

Projection to end of inspection interval must assure that hold-down force maintained through next inspection interval.

Approach: Value determined by plant-specific design requirements.

WCAP- 17096-NP December 2009 Revision 2

E-28 W-ID: 8 Thermal Shield Assembly Thermal Shield Flexures Category: Primary Applicability: All plants with thermal shields Degradation Effect: Cracking (fatigue)

Expansion Link: None Function: The flexure is the lower structural support for the thermal shield. Flexures hold the thermal shield concentric to the core. The flexure design allows for differential thermal expansion between the core barrel and the thermal shield.

Inspection Method: Visual (VT-3) no later than 2 refueling outages from the beginning of the license renewal period. Subsequent examinations on a 10-year interval.

Coverage: 100% of thermal shield flexures See MRP-227 Figures 4-29 and 4-36.

Observable Effect: Crack, displacement, fracture, or component separation.

Failure along weld at base of flexure or failure of weld attachment to thermal shield.

Failure Failure Mechanism: Large deflections of the flexure due to thermalcycling may lead to fatigue failures.

Failure Effect: Failure of flexures contributes to vibration of the thermal shield. Failure can also result in flow blockage, wear, and damage to specimen guides.

Failure Criteria: Number of unfailed thermal shield flexures must be sufficient to retain structural functionality of the entire thermal shield assembly.

Methodology Goal: Determine the number and location of thermal shield flexures that must remain intact to retain structural functionality of the entire thermal shield assembly.

Data Requirements:

• Load

• Geometry

• History (transients)

• Materials Analysis: Perform structural assessment to determine the minimum number of flexures required to retain structural integrity.

The dynamic response of the thermal shield should be established.

Assume:

1. Any thermal shield flexure with an observed flaw has failed.

2. No credit for "bumpers" and other redundant structures.

Acceptance Criteria: Failure of a thermal shield flexure is acceptable if it can be demonstrated that the dynamic response of the thermal shield is unchanged when the flexure is removed from the model.

Any observation of a failed thermal shield flexure should lead to enhanced vigilance for fatigue and vibration monitoring systems.

Approach: Plant-specific analysis.

WCAP- 17096-NP December 2009 Revision 2

F-1 APPENDIX F FLOW CHARTS OF ILLUSTRATING EVALUATION METHODOLOGIES FOR WESTINGHOUSE-DESIGNED PLANTS Westinghouse Primary and Expansion Components W-ID: 1 Control Rod Guide Tube Assembly - Guide Pates (Cards)

W-ID: 2 Control Rod Guide Tube Assembly - Lower Flange Welds W-ID: 2.1 Lower Support Assembly - Lower Support Column Bodies (Cast)

W4D: 2.2 Bottom Mounted Instrumentation (BMI) System - BMI Column Bodies W-ID: 3 Core Barrel Assembly - Upper Core Barrel Flange Weld W-ID: 3.1 Core Barrel Assembly - Core Barrel Flange, Core Barrel Outlet Nozzles, Lower Core Barrel Flange Weld W-1D: 3.2 Lower Support Assembly - Lower Support Columns (non cast)

W-ID: 4 Baffle-Former Assembly - Baffle-Edge Bolts W-ID: 5 Baffle-Former Assembly - Baffle-Former Bolts W-ID: 5.1 Core Barrel Assembly - Barrel-Former Bolts W-ID: 5.2 Lower Support Assembly - Lower Support Column Bolts W-ID: 6 Baffle-Former Assembly - Assembly W-ID: 7 Alignment and Interfacing Components - Internal Hold-down Spring W-ID: 8 Thermal Sleeve Assembly - Thermal Shield Flexures WCAP- 17096-NP December 2009 Revision 2

F-2 W-ID:1 Control Rod Guide Tube Assembly Guide Plates (Cards)

Wear Volume Linworn Stage A Stage B WCAP- 17096-NP December 2009 Revision 2

F-3 W-ID: 2 Control Rod Guide Tube Assembly Lower Flange Welds of Control Rod Guide Tube VVT-1 raklefawi No Arron+

LowerwelFlange .

/_W-ID: 2.1 WedsBunedy Mitigate1 VT-3 of Castl WI: .

Lower Support VT-3 of BMI Acceptable Weldl Column Column Bodies Patterns /

Components Minimum Weld Analysis Yes Determine Guide Tube Assembly Select Candidate Deflection with Evaluate: 4 Pattern >Njo ýTable of Rejected Pattern of Failed Failed Welds --- o Allowable deflection Acceptable'? Weld Patterns Welds Removed from from Design Analysis Analysis Select Limiting Faulted Condition WCAP- 17096-NP December 2009 Revision 2

F-4 W-ID: 2.1 Lower Support Structure Lower Support Column Bodies (Cast)

December 2009 WCAP-17096-NP WCAP- 17096-NP December 2009 Revision 2

F-5 W-ID:2.2 Bottom Mounted Instrumentation BMI Column Bodies WCAP- 17096-NP December 2009 Revision 2

F-6 W-ID:3 Core Barrel Assembly Upper Core Barrel Flange Welds Option 1 OD Inspection Yes No Determine applied stres, intensity factor, Kiapp, for projected length under faulted condition Note: Flaw dimension Lc is limit where growth prohibited by stress distribution.

WCAP- 17096-NP December 2009 Revision 2

F-7 W-ID: 3 Core Barrel Assembly Upper Core Barrel Flange Weld Option 2: ID Inspection No Note: Flaw dimension Lc is limit where growth is prohibited by stress distribution WCAP- 17096-NP December 2009 Revision 2

F-8 W-ID: 3.1 Core Barrel Assembly Core Barrel Flange, Core Barrel Outlet Nozzles and Lower Core Barrel Flange Welds 7Crack-like indication in" Core Barrel to Lower Core Plate Weld z "rack ýGrothRate No Yes Expansion

>2" Estimate Lower Bound Fracture Toughness, Klc, at Flaw Location WCAP- 17096-NP December 2009 Revision 2

F-9 W-ID: 3.2 Lower Support Assembly Lower Support Column Bodies (Non Cast)

WCAP- 17096-NP December 2009 Revision 2

F-10 W-ID: 4 Baffle-Former Assembly Baffle Edge Bolts FMEA Layout for Bolts ACTIVITY KEY INPUTS DISPOSITION HistoicalRecord and

2. Missing *I

3. Broken

4. Improperly Installed Devlo Jutfiaio o Develop Justification for Continued Operation IU II EdeBl Historical * *Record DcmnandS SOff-Normall *1. Protruding Hea d _______ istoicalDocument Conditions 2 . Missing*

3**.Brokeng

,4. Improperly Installed Develop Justification for Continued Operation rS W-ID: 4 Edge Bolts CHistorical Record and and l l~~~~.

Protruding Head Historica " Document Douetf S Genera;1 2. Missing Genera 3. Broken Visult ,4. Improperly Installed -'1 Inspection Visa NwCniudOperation Develop Justification for rn Continued Operation _

Normal Recoand Conditions ,Dcmn WCAP- 17096-NP December 2009 Revision 2

F-Il W-1D: 5 Baffle-Former Assembly Baffle-Former Bolts WCAP- 17096-NP December 2009 Revision 2

F-12 W-ID: 5.1 Baffle-Former Assembly Barrel-Former Bolts WCAP- 17096-NP December 2009 Revision 2

F- 13 W-ID:: 5.2 Lower Support Assembly Lower Support Column Bolts WCAP- 17096-NP December 2009 Revision 2

F-14 W-ID: 6 Baffle-Former Assembly FMEA Layout KEYINPUTS December 2009 WCAP-WCAP- 17096-NP December 2009 Revision 2

F-15 W-ID: 7 Alignment and Interfacing Devices Internals Hold Down Spring.

W-ID: 7 Measurement of Hold Down Spring Height Hold Down Force and Spring Height Specifications WCAP- 17096-NP December 2009 Revision 2

F-16 W-ID: 8 Thermal Shield Assembly Thermal Shield Flexures WCAP- 17096-NP December 2009 Revision 2